Mychailo B. Toloczko

Direct detection of Pb in urine and Cd, Pb, Cu, and Ag in natural waters using electrochemical sensors immobilized with DMSA functionalized magnetic nanoparticles

Urine is universally recognized as one of the best non-invasive matrices for biomonitoring exposure to a broad range of xenobiotics, including toxic metals. Detection of metal ions in urine has been problematic due to the protein competition and electrode fouling. For direct, simple, and field-deployable monitoring of urinary Pb, electrochemical sensors employing superparamagnetic iron oxide (Fe3O4) nanoparticles with a surface functionalization of dimercaptosuccinic acid (DMSA) has been developed. The metal detection involves rapid collection of dispersed metal-bound nanoparticles from a sample solution at a magnetic or electromagnetic electrode, followed by the stripping voltammetry of the metal in acidic medium. The sensors were evaluated as a function of solution pH, the binding affinity of Pb to DMSA-Fe3O4, the ratio of nanoparticles per sample volume, preconcentration time, and Pb concentrations. The effect of binding competitions between the DMSA-Fe3O4 and urine constituents for Pb on the sensor responses was studied. After 90 s of preconcentration in samples containing 25 vol.% of rat urine and 0.1 g L(-1) of DMSA-Fe3O4, the sensor could detect background level of Pb (0.5 ppb) and yielded linear responses from 0 to 50 ppb of Pb, excellent reproducibility (%RSD of 5.3 for seven measurements of 30 ppb Pb), and Pb concentrations comparable to those measured by ICP-MS. The sensor could also simultaneously detect background levels (<1 ppb) of Cd, Pb, Cu, and Ag in river and seawater.

Irradiation creep and void swelling of austenitic stainless steels at low displacement rates in light water energy systems

Abstract The majority of the high fluence data on the void swelling and irradiation creep of austenitic steels was generated at relatively high displacement rates characteristic of fast reactors. The application of these data to other reactor systems which operate at lower displacement rates requires that some estimate be made of the dependence of swelling and irradiation creep on displacement rate. Austenitic components in typical light water reactors (LWRs) experience displacement rates that are an order of magnitude or more lower than that found in fast reactors. Since irradiation creep has a component that is proportional to swelling, it is anticipated that irradiation creep will exhibit a sensitivity to displacement rate that is a direct expression of the dependence of swelling on displacement rate. Until recently, however, the non-swelling-related component of creep was also thought to exhibit its own dependence on displacement rate, increasing at lower fluxes. This perception originally arose from the work of Lewthwaite and Mosedale on irradiation creep of cold-worked steels in the Dounreay Fast Reactor at temperatures in the 270–350°C range. It now appears that this interpretation is incorrect and, in fact, the steady-state value of the non-swelling component of irradiation creep is relatively insensitive to displacement rate. The earlier perceived flux dependence appears to have arisen from a failure to properly separate the transient and steady-state regimes of irradiation creep. As evidence begins to accumulate on void swelling of stainless steels at LWR-relevant displacement rates, voids are being observed at lower-than-expected temperatures and often at very low dpa levels, implying that the ‘temperature shift’ phenomenon may indeed be operating to increase swelling. It is demonstrated in this paper that if swelling exceeds ∼ 10% at PWR-relevant temperatures, stainless steels become very brittle.

Ductility correlations between shear punch and uniaxial tensile test data

The shear punch test was developed to address the need of the fusion reactor structural materials community for small scale mechanical properties tests. It has been demonstrated that effective shear strength data obtained from the shear punch test can be linearly related to uniaxial tensile strength for a wide variety of alloys. The current work explores the existence of a similar relationship between shear punch test data and both the tensile strain hardening exponent and the uniform elongation.

Shear punch tests performed using a new low compliance test fixture

Based on a recent finite element analysis (FEA) study performed on the shear punch test technique, it was suggested that compliance in a test frame and fixturing which is quite acceptable for uniaxial tensile tests, is much too large for shear punch tests. The FEA study suggested that this relatively large compliance was masking both the true yield point and the shape of the load versus displacement trace obtained in shear punch tests. The knowledge gained from the FEA study was used to design a new shear punch test fixture which more directly measures punch tip displacement. The design of this fixture, the traces obtained from this fixture, and the correlation between uniaxial yield stress and shear yield stress obtained using this fixture are presented here. In general, traces obtained from the new fixture contain much less compliance resulting in a trace shape which is more similar in appearance to a corresponding uniaxial tensile trace. Due to the more direct measurement of displacement, it was possible to measure yield stress at an offset shear strain in a manner analogous to yield stress measurement in a uniaxial tensile test. The correlation between shear yield and uniaxial yield was altered by this new yield measurement technique, but the new correlation was not as greatly improved as was suggested would occur from the FEA study.

Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold-Rolled Alloy 690 in Pressurized Water Reactor Primary Water

Grain boundary microstructures and microchemistries are examined in cold-rolled Alloy 690 (UNS N06690) materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in pressurized water reactor (PWR) primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill-annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. For the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results su...

High Temperature Tensile Testing of Modified 9Cr-1Mo after Irradiation with High Energy Protons

Abstract This study examines the effect of tensile test temperatures ranging from 50 to 600 °C on the tensile properties of a modified 9Cr–1Mo ferritic steel after high energy proton irradiation at about 35–67 °C to doses from 1 to 3 dpa and 9 dpa. For the specimens irradiated to doses between 1 and 3 dpa, it was observed that the yield strength and ultimate strength decreased monotonically as a function of tensile test temperature, whereas the uniform elongation (UE) remained at approximately 1% for tensile test temperatures up to 250 °C and then increased for tensile test temperatures up to and including 500 °C. At 600 °C, the UE was observed to be less than the values at 400 and 500 °C. UE of the irradiated material tensile tested at 400–600 °C was observed to be greater than the values for the unirradiated material at the same temperatures. Tensile tests on the 9 dpa specimens followed similar trends.

Effect of low temperature irradiation on the mechanical properties of ternary V–Cr–Ti alloys as determined by tensile tests and shear punch tests

Abstract Tensile tests and shear punch tests were performed on a variety of vanadium alloys that were irradiated in the Advanced Test Reactor (ATR) at temperatures between 200°C and 300°C to doses between 3 and 5 dpa. Tests were performed at room temperature and the irradiation temperature. The results of both the tensile tests and the shear punch tests show that following low temperature irradiation, the yield strength (YS) increased by a factor of 3–4 while the ultimate strength increased by a factor of approximately 3. Uniform elongation (UE) and tensile reduction in area show that the ductility diminishes following irradiation. The correlation between uniaxial ultimate strength and effective shear maximum strength was in excellent agreement with previous studies on other materials. Using the room temperature test data, the correlation between uniaxial YS and effective shear YS was in excellent agreement with previous studies on other materials.

Comparison of fission neutron and proton/spallation neutron irradiation effects on the tensile behavior of type 316 and 304 stainless steel

Abstract As part of the accelerator production of tritium and the spallation neutron source programs, the tensile properties of annealed 304L, 316LN and 316L stainless steel have been measured after proton and spallation neutron irradiation in the target region of the Los Alamos Neutron Science Center (LANSCE) accelerator (800 MeV, 1 mA) to a maximum dose of 12 dpa at temperatures ranging from 30 to 120 °C. In addition to the displacement damage produced from the irradiation, up to several thousand atomic parts per million (appm) of hydrogen and helium were produced in the irradiated material via spallation reactions. Results of tensile tests at temperatures from room temperature up to 164 °C show large increases in tensile yield strength, small increases in ultimate tensile strength, reductions in strain hardening capacity and reductions in ductility (uniform elongation and strain-to-necking) with increasing irradiation dose. A comparison of these data with the large database on tensile properties of type 316 stainless steel exposed to fission neutrons and tensile tested over the same temperature range show similar trends with regard to strength changes, but significantly larger reductions in ductility with irradiation dose were observed after irradiation in the spallation environment. The much higher amounts of helium and hydrogen produced through spallation in the LANSCE spectrum, compared to those developed in fission neutron irradiation environments at equivalent doses, may contribute to degradation in ductility.

Irradiation creep and void swelling of two LMR heats of HT9 at ∼ 400°C and 165 dpa☆

Abstract Two nominally identical heats of HT9 ferritic—martensitic steel were produced, fabricated into pressurized tubes, and then irradiated in FFTF, using identical procedures. After reaching 165 dpa at ∼ 400°C, small differences in strains associated with both phase-related changes in lattice parameter and void swelling were observed in comparing the two heats. The creep strains, while different, exhibited the same functional dependence on swelling behavior. The derived creep coefficients, the one associated with creep in the absence of swelling and the one directly responsive to swelling, were essentially identical for the two heats. Even more significantly, the creep coefficients for this bcc ferritic-martensitic steel appear to be very similar and possibly identical to those routinely derived from creep experiments on fcc austenitic steels.

One Dimensional Cold Rolling Effects on Stress Corrosion Crack Growth in Alloy 690 Tubing and Plate Materials

Stress corrosion crack-growth experiments have been performed on cold-rolled alloy 690 materials in simulated PWR primary water at 360°C. Extruded alloy 690 CRDM tubing in two conditions, thermally treated (TT) and solution annealed (SA), was cold rolled (CR) in one direction to several reductions reaching a maximum of 31% and tested in the S-L orientation. High stress corrosion cracking (SCC) propagation rates (~8x10−8 mm/s) were observed for the 31%CR alloy 690TT material, while the 31%CR alloy 690SA exhibited 10X lower rates. The difference in intergranular SCC susceptibility appears to be related to grain boundary carbide distribution before cold rolling. SCC growth rates were found to depend on test temperature and hydrogen concentration. Tests were also performed on two alloy 690 plate heats, one CR to a reduction of 26% and the other to 20%. SCC growth rates at 360°C were similar to that measured for the 31%CR alloy 690TT CRDM tubing. Comparisons will be made to other results on CR alloy 690 materials.