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Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with 50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520–620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350–650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

Oxidation experiments were conducted at 1000–1200 °C in flowing steam with samples of as-received Zr-1Nb alloy E110 tubing and/or polished E110 tubing. The purpose was to determine the oxidation behavior of this alloy under postulated loss-of-coolant accident conditions in light water reactors. The as-received E110 tubing exhibited a high degree of susceptibility to nodular oxidation and breakaway oxidation at relatively low test times, as compared to other cladding alloys. The nodules grew much more rapidly at 1000 °C than 1100 °C, as did the associated hydrogen uptake. The oxidation behavior was strongly affected by the surface condition of the materials. Polishing to ≈0.1 μm roughness (the roughness of the as-received tubing was ≈0.4 μm) delayed breakaway oxidation. Polishing also removed surface impurities. For polished samples oxidized at 1100 °C, no significant nodular oxidation appeared up to 1000 s. For polished samples oxidized at 1000 °C, hydrogen uptake >100 wppm was delayed from ≈300 s to >900 s. Weight-gain coefficients were determined for pre-breakaway oxidation of polished-only and machined-and-polished E110 tubing samples: 0.162 (mg/cm 2 )/s 0.5 at 1000 °C and 0.613 (mg/cm 2 )/s 0.5 at 1100 °C.

The article reports the results of experimental campaigns on plasma ohmic heating performed during 1999-2000 on the spherical tokamak Globus-M. Later experimental results with the tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field and plasma pulse duration in these experiments were significantly increased. The method of stray magnetic field compensation is described. The technology of vacuum vessel conditioning, including boronization of the vessel performed at the end of the experiments, is briefly discussed. Specific features of neutral gas column breakdown in spherical tokamaks by applied inductive voltage are outlined. Also discussed is the influence of ECR preionization on the breakdown conditions. Experimental data on plasma column formation and current ramp-up in different regimes of operation with the magnetic flux of the central solenoid limited to ~100 mWb are presented. A significant reduction of the background density after boronization (below 2 × 1018m-3) allows the density to be completely controlled with external gas puffing and makes the influence of the wall negligible. The magnetic flux consumption efficiency is discussed. The results of magnetic equilibrium simulations are presented and compared with experiment. Ramp-up of the plasma current of 0.25 MA for a time interval of about 0.03 s with about 0.02 s flat-top at a toroidal field strength of 0.35 T allows the conclusion that the target design parameters of Globus-M could be achieved in a double swing regime.

Test results for actively cooled, graphite RGTi type brazed to molybdenum tube divertor mock-ups are presented. Particular attention was paid to finding a reliable, non-destructive method of investigating brazed joints. The mock-ups were ultrasonically tested and the results correlated with the microstructure of the joints. Two types of tests were carried out: screening tests with a stepped increase of the surface heat flux from shot to shot and thermal cycling tests at a fixed power density level. The heat loads were varied in the range from 6 MW/m2 to 12 MW/m2. About 150 cycles of heat loads with a power density of 10 MW/m2 were realized and no macrodamage was found. The heat flux experiments were accompanied by detailed metallographic investigations.

Different materials (W, Mo, Ta, Ta + W and carbon-based materials) have been tested with respect to their damage behaviour (thermal erosion, melting, cracks formation) at conditions simulating plasma disruption. A quasistationary plasma accelerator (pulse duration of 0.1 ms and incident energy density of 5–12 MJ/m 2 ) and a pulse electron accelerator (pulse duration of 0.05–0.2 ms and incident energy density of 1–5 MJ/m 2 ) have been used for simulation of the thermal quench phase of plasma disruption. It was shown that for metals the absorbed energy is only 10% from the total one. The thermal erosion by plasma irradiation is many times lower than the predicted one. However under electron beam shocks the thermal erosion is comparable with the theory. A microstructural study has shown that tungsten exhibits the lowest crack formation resistance, on the other hand cracks in Ta and Ta + W alloy have not been observed. The thermal erosion of the tested carbon base materials in these experiments is smaller than predicted.