Suresh K. Yagnik

Round-Robin Testing of Fracture Toughness Characteristics of Thin-Walled Tubing

Cladding fracture behavior is an important consideration, particularly in secondary damage of fuel cladding during service and during handling and storage of discharged fuel. A number of test techn ...

Ductility of Zircaloy-4 Fuel Cladding and Guide Tubes at High Fluences

Zircaloy fuel cladding suffers progressive degradation of ductility as its neutron exposure and hydrogen uptake increase with burnup. The loss of ductility appears to be the key property governing the cladding integrity in service. We report ductility data of Zircaloy-4 fabricated in stress-relief annealed (SRA) and recrystallized (RXA) conditions, covering a range of fluence, hydrogen content, and irradiation and test temperature. The testing was performed on unirradiated and irradiated Zircaloy-4 cladding and guide tubes in SRA and RXA conditions, respectively, in the temperature range 25–350°C. These materials had been exposed to an estimated neutron fluence of ∼8 to 10 × 1025 n/m2 (E > 1 MeV) over four cycles of PWR operation. Due to its RXA fabrication condition, the guide tube material was also believed to represent BWR cladding. The hydrogen contents in the irradiated cladding was in the range of ∼200–600 ppm, exhibiting typical radial distribution of circumferential hydrides. By comparison, the hydrogen content in the irradiated guide tube material was in the range of ∼250–1800 ppm, exhibiting fairly uniform through-wall distribution of circumferential hydrides. Tensile tests and hydraulic burst tests were conducted on 130–150 mm long tubular specimens. In addition, smaller specimens machined in the form of 55 mm long curvilinear dog-bone and 10 mm slotted semicircular arc were tested in plane stress and plane strain configurations. Corresponding unirradiated archive materials in as-received condition and with uniform hydrogen charging up to 1200 ppm were also tested by identical methods. In all tests the fracture mode was examined by SEM fractography. The investigations revealed a decrease in ductility of Zircaloy-4, mainly caused by irradiation and only partly by increasing hydrogen content. Unlike the elongation data, the strength data remained nearly constant with increasing hydrogen content in both materials at all test temperatures. The irradiated RXA material showed better ductility than irradiated SRA material at equivalent hydrogen levels, and exhibited a clearer correlation of ductility with hydrogen content, mainly due to its uniform hydrogen distribution. The paper will provide these and other quantitative data, e.g., those correlating ductility with the local (near fracture surface) hydrogen content. The paper synthesizes the experimental results and discusses their possible application to the criteria for hydrogen concentration and ductility limits in high burnup fuel.

Steam oxidation of fuel in defective LWR rods

Abstract Oxidation of UO 2 by pure steam at pressures of 7 and 70 atm and 500°C and 600°C was measured in a thermogravimetric apparatus. The kinetics are linear, vary as the square root of the steam pressure, and are consistent with initial rates extrapolated from higher-temperature experiments in 1-atm steam. At temperatures characteristic of normal operation of defective fuel rods, the rate of hydrogen production by thermal oxidation of the fuel in steam is small compared with that due to cladding corrosion. The presence of H 2 in the steam has a much greater retarding influence on fuel oxidation than on cladding oxidation. Other potential sources of fuel chemical reactivity in steam, including reaction in cracks in the hot pellet interior and radiolysis of steam by recoiling fission fragments, do not result in significant fuel oxidation. During the incubation stage of fuel-rod degradation, the bulk of the evidence indicates that fuel oxidation is not a major source of the hydrogen in the fuel–cladding gap that eventually may cause secondary-hydriding failure of the rod.

High pressure hydriding of sponge-Zr in steam-hydrogen mixtures

Abstract Hydriding kinetics of thin sponge-Zr layers metallurgically bonded to a Zircaloy disk has been studied by thermogravimetry in the temperature range 350–400°C in 7 MPa hydrogen-steam mixtures. Some specimens were prefilmed with a thin oxide layer prior to exposure to the reactant gas; all were coated with a thin layer of gold to avoid premature reaction at edges. Two types of hydriding were observed in prefilmed specimens, viz., a slow hydrogen absorption process that precedes an accelerated (massive) hydriding. At 7 MPa total pressure, the critical ratio of H2/H2O above which massive hydriding occurs at 400°C is ∼ 200. The critical H2/H20 ratio is shifted to ∼2.5 × 103 at 350°C. The slow hydriding process occurs only when conditions for hydriding and oxidation are approximately equally favorable. Based on maximum weight gain, the specimen is completely converted to δ-ZrH2 by massive hydriding in ∼5 h at a hydriding rate of ∼10−6 mol H/cm2 s. Incubation times of 10–20 h prior to the onset of massive hydriding increases with prefilm oxide thickness in the range of 0–10 μm. By changing to a steam-enriched gas, massive hydriding that initially started in a steam-starved condition was arrested by re-formation of a protective oxide scale.

High pressure oxidation of sponge-Zr in steam/hydrogen mixtures

Abstract A thermogravimetric apparatus for operation in 1 and 70 atm steam-hydrogen or steam-helium mixtures was used to investigate the oxidation kinetics of sponge-Zr containing 215 ppm Fe. Weight-gain rates, reflecting both oxygen and hydrogen uptake, were measured in the temperature range 350–400°C. The specimens consisted of thin sponge-Zr layers metallurgically bonded to a Zircaloy disk. The edges of the disk specimens were coated with a thin layer of pure gold to avoid the deleterious effect of corners. Following each experiment, the specimens were examined metallographically to reveal the morphology of the oxide and/or hydride formed. Two types of oxide, one black and uniform and the other white and nodular, were observed on sponge-Zr surfaces oxidized in steam environments at 70 atm. The oxidation rate when white-nodular oxide formed was a factor of two higher than that of black-uniform oxide at 400°C for steam contents above 1 mol%. The oxidation rate was independent of total pressure, the carrier gas (H 2 or He) and steam content above ∼1 mol%. The oxidation kinetics of sponge-Zr follows a linear law for maximum reaction times up to ∼ 6 days. The oxidation rate in steam-hydrogen mixtures at 70 atm total pressure decreases when the steam content approaches the steam-starved region (∼ 0.5 mol% steam at 400°C and ∼ 0.02 mol% steam at 350°C). Lower steam concentrations cause massive hydriding of the specimens. Even at steam concentrations above the critical value, direct hydrogen absorption from the gas was manifest by hydrogen pickup fractions greater than unity.

Effect of Local Hydride Accumulations on Zircaloy Cladding Mechanical Properties

Mechanical response of fuel cladding with local hydride accumulations is crucial in the assessment of cladding integrity at high burn-ups. We have performed high-temperature low-strain rate burst tests on irradiated cladding samples with and without hydride lenses or blisters to seek answers to the following questions: Does the presence of a hydride lens inevitably lead to rupture at a lower pressure? How does it mechanistically affect the crack initiation and propagation? The irradiated samples in our investigation were taken from the regions of the fuel cladding with oxide spallation. Subsequently, we used neutron radiography to further select samples covering a range of hydride blister sizes on which the burst testing was performed. Rupture pressure, hoop strength, and circumferential strain data will be reported. For each sample tested, detailed metallography and fractography were performed on 2-mm size sections containing the burst opening to provide insights into the mechanism of crack initiation and propagation. Local and mean hydrogen concentrations were measured. The paper will include and elucidate new details often not fully investigated by other burst test investigations reported in the open literature. In samples with multiple blisters, the crack initiates at the largest one, which also governs the fracture mode. Reduction in the rupture pressure can be simply correlated to the reduction in sample wall thickness excluding the blister (i.e., its remaining ligament). There is a lower bound on the blister size to have any influence on the rupture pressure. Further, local plastic circumferential strain at each blister can be correlated to relative hydride lens area, as projected onto the cladding surface.

Chemical processes in defective LWR fuel rods

Abstract The results of several experimental studies aimed at improving understanding of the chemical processes that cause severe degradation of defective light-water reactor fuel cladding are reported. The competition between oxidation and hydriding of zirconium and zircaloy exposed to steam-hydrogen mixtures at 70 bar and 350–400°C was studied by thermogravimetry. A critical H 2 /H 2 O ratio of the gas separates regimes of these two types of reaction. For sponge-Zr, the critical ratios at 350 and 400°C are ≈ 5000 and ≈ 200, respectively. Radiolysis of steam by alpha particles was studied mass spectrometrically. The yield of the hydrogen radiolysis product in saturated steam at 290°C was found to be ≈ 8 molecules per 100 eV of deposited energy. An in-reactor experiment demonstrated that fission-fragment-irradiated steam can oxidize UO 2 to UO 2+ x .

High-pressure hydriding of Zircaloy

Abstract The hydriding characteristics of Zircaloy-2(Zry), sponge zirconium (as a liner on Zry plate), and crystal-bar zirconium exposed to pure H2 at 0.1 MPa or 7 MPa and 400°C were determined in a thermogravimetric apparatus. The morphology of the hydrided specimens was also examined by optical microscopy. For all specimen types, the rate of hydriding in 7 MPa H 2 was two orders of magnitude greater than in 0.1 MPa H 2 . For Zry, uniform bulk hydriding was revealed by hydride precipitates at room temperature and on one occasion, a sunburst hydride. In addition, all specimen types exhibited a hydride surface layer. In a duplex Zry/sponge-Zr specimen, Zry is more heavily hydrided than the sponge Zr layer.

Effect of Hydrides on Mechanical Properties and Failure Morphology of BWR Fuel Cladding at Very High Strain Rate

The data on the cladding mechanical response under rapid deformation are crucial for the safety assessment of LWR fuel rods for reactivity initiated accident (RIA) and for spent fuel shipping and storage cask-drop accidents. The reported data on the mechanical properties of the irradiated cladding under high strain rates, however, are very limited, and conventional axial tensile data measured at slow strain rates may not be directly applicable to RIA or drop impact analyses. The objective of this work was to obtain basic mechanical properties such as the stress-strain relationship, ductility, and the critical strain energy density (SED) of fuel cladding with various hydride morphologies at high deformation rate. A unique rapid burst test method was employed on open-end tube specimens, achieving a strain rate of ∼1 s−1 and specimen rupture in ∼10 ms. Unirradiated and irradiated Zry-2 fuel cladding specimens were tested from room temperature up to 350°C. The mechanical properties data were assessed in terms of SED and failure strain, and failure morphologies were examined by optical and scanning electron microscopy. The tests performed at room temperature on unirradiated cladding with circumferential hydrides revealed that the failure elongation was unchanged or increased slightly with an increase the in hydrogen content up to ∼100 wt ppm but decreased drastically at ∼400 wt ppm to values approaching 0 %. A drastic reduction in failure strain occurred at a lower hydrogen concentration of ∼100 wt ppm for radial hydrides, compared to ∼200–400 wt ppm for the circumferential hydride distribution. The fraction of specimen thickness occupied by the accumulated length of hydrides in radial direction was a better predictive indicator of specimen failure and ductility reduction than the total hydrogen content. In addition, the high strain rates did not seem to seriously impact stress-strain behavior when hydrogen content is >400 wt ppm. The data analyses revealed smaller values of the strain hardening exponent (n) compared to those from the conventional data for the slow strain rates, indicating that the plastic instability theory simply is not appropriate to evaluate the failure strain under RIA conditions.

Experimental Investigation of Irradiation Creep and Growth of Recrystallized Zircaloy-4 Guide Tubes Pre-Irradiated in PWR

Re-crystallized Zircaloy-4 guide tubes were irradiated in commercial pressurized water reactors (PWRs) at three different temperatures to fluences near 1×1022 n/cm2 E>1 MeV (15 displacements per atom), resulting in moderate corrosion and three different hydrogen contents (approximately 135, 240, and 700 parts per million). Sections of the guide tubes were re-irradiated in the Halden reactor to assess the irradiation creep and growth behaviors. Three conditions were applied: Bellows-loaded axial compression creep; zero stress growth; and zero stress, zero flux (control specimens). The guide tube sections were re-irradiated under simulated PWR conditions by utilizing a pressurized light water loop operating with normal PWR water chemistry at approximately 320°C. Axial length changes were measured in-reactor by linear variable differential transformers (LVDTs), and post-irradiation hot cell measurements were done to confirm the LVDT elongation measurements. After minor corrections were made to account for reactor testing variables, it was shown that the LVDT measurements were accurate, thus creep and growth or free growth rates were established for each guide tube section. Hot cell examinations were also performed to establish the state of corrosion of each specimen, including hydrogen content, both before and after the re-irradiation in the Halden reactor. The results showed that stress free irradiation growth was different for each specimen and correlated qualitatively with the hydrogen content and commercial irradiation temperatures of the guide tubes. The higher hydrogen content, or higher commercial irradiation temperature, gave rise to higher subsequent growth rates. After subtracting the growth strain from the measured creep and growth strain values, the magnitude of creep and creep rates were essentially the same for all specimens: No effect of commercial reactor irradiation temperature or hydrogen content was observed. The results give important new data on irradiation creep and growth and on the correlation between hydrogen content and irradiation temperature on growth rates.