The Idaho National Laboratory is conducting moderate strain rate (5 to 200 per second) research on stainless steel materials in support of the Department of Energy's National Spent Nuclear Fuel Program. For this research, strain rate effects are characterized by comparison to quasi-static tensile test results. Considerable tensile testing has been conducted resulting in the generation of a large amount of basic material data expressed as engineering and true stress-strain curves. The purpose of this paper is to present the results of quasi-static tensile testing of 304L and 316L stainless steels in order to add to the existing data pool for these materials and make the data more readily available to other researchers, engineers, and interested parties.Standard tensile testing of round specimens in accordance with ASTM procedure A 370-03a was conducted on 304L and 316L stainless steel plate materials at temperatures ranging from -20°F to 600°F. Two plate thicknesses, eight material heats, and both base and weld metal were tested. Material yield strength, ultimate strength, ultimate strain, fracture strength, fracture strain and reduction in area were determined. Engineering and true stress-strain curves to failure were developed and comparisons to ASME Code minimums were made. The procedures used during testing and the typical results obtained are presented in this paper. INTRODUCTIONThe Department of Energy's (DOE) National Spent Nuclear Fuel Program (NSNFP), working with the Office of Civilian Radioactive Waste Management (OCRWM), the Idaho National Laboratory (INL) and other DOE sites, has supported development of canisters for loading and interim storage, transportation, and disposal of DOE spent nuclear fuel (SNF). To assess the integrity of these SNF canisters under dynamic, impact loading, the INL is conducting moderate strain rate (5 to 200 per second) research on 304L and 316L stainless steels which are the preferred materials for construction. The goal of this research is to define and justify elevated strain rate effects for these materials over a range of applicable temperatures and develop corresponding true stress-strain relationships that can be used to perform accurate analytical assessments of canister impact events. Both base metal and weld metal are of significance and are being investigated.
Stainless steels are used for the construction of numerous spent nuclear fuel or radioactive material containers that may be subjected to high strains and moderate strain rates (10 to 200 per second) during accidental drop events. Mechanical characteristics of these materials under dynamic (impact) loads in the strain rate range of concern are not well documented. The goal of the work presented in this paper was to improve understanding of moderate strain rate phenomena on these materials. Utilizing a drop-weight impact test machine and relatively large test specimens (1/2-inch thick), initial test efforts focused on the tensile behavior of specific stainless steel materials during impact loading.Impact tests of 304L and 316L stainless steel test specimens at two different strain rates, 25 per second (304L and 316L material) and 50 per second (304L material) were performed for comparison to their quasi-static tensile test properties. Elevated strain rate stress-strain curves for the two materials were determined using the impact test machine and a "total impact energy" approach. This approach considered the deformation energy required to strain the specimens at a given strain rate. The material data developed was then utilized in analytical simulations to validate the final elevated stress-strain curves. The procedures used during testing and the results obtained are described in this paper.
The Idaho National Engineering and Environmental Laboratory (INEEL) developed an apparatus capable of supporting a wide variety of material studies and distinct component testing under impact loads. Material studies include material (metals, plastics, concrete, etc.) response due to bending, tension, shear, and compression loadings at elevated strain rates. Similar testing can also be performed on any distinct component fitting within the apparatus impact loading volume. This apparatus is referred to as the Impact Test Machine (ITM). The ITM is initially being used by the Department of Energy (DOE) to test 304L and 316L stainless steel tensile test specimens at various strain rates for comparison to static properties. The goal is to ultimately develop true stress-strain curves at various strain rates and temperatures for these steels. These curves can then be used in analytical simulations to more accurately predict the deformation and resulting material straining in spent nuclear fuel (SNF) containers, canisters, and casks under accidental drop events (Ref: Snow 1999, 2000). Test results can also help determine a basis for establishing allowable strain limits for these large deformation, inelastic events. This material investigation is currently in an early stage of development. This paper will discuss the results of tensile tests performed on test specimens employed in the formulation of the test process and initial checkout of the ITM.
(702)-895-4328 / FAX (7021-895-3936 ABSTRACT Linear and non-linear stresses and buckling loads in the pintle, top head, canister shell and top plate of two high-level spent nuclear fuel canister models are evaluated under a variety of static and dynamic loading conditions resulting from normal handling. The proposed canister designs are made of 304L stainless steel, having a yield strength of 220 Mpa (32000 p s i) , an ultimate tensile strength of 586 Mpa (85000 psi) and a high resistance to crack initiation. The canisters are 0.0095m (0.375 in) thick, 0.7m (28 in) in diameter and 4.2m (165 in) long. Finite element analysis is used in the stress and stability analysis. Four node thin shell elements and 3-D isoparametric four node thick shell elements are used in modeling the weldment, tophead, pintle, bottom plate and the canister shell. Results obtained from the various types of analysis and the two finite element models, compare favorably within the limitation and applicability of each element type, Results are also compared with theory for some canister loading conditions when a theoretical analysis is possible. Results show that these proposed container models are not adequate to handle stresses and some local and global instabilities resulting from accidental drops, impact between canisters and sudden slippage of the crane cable, during the process of lifting and handling of the nuclear fuel containers. ABSTRACT The thickness and time at failure oithe i O O m m thck overpack and the 9.5mm thick inner container of a Multi purpose canister have been assessed due to loads resulting fiom temperature, overburden, bacldill pressure and seismic loads. Critical stresses at various reduced thicknesses, resuiting h m pitting corrosion over the years of empiacement,
The National Spent Nuclear Fuel Program (NSNFP) at the Idaho National Engineering and Environmental Laboratory (INEEL) prepared four representative Department of Energy (DOE) spent nuclear fuel (SNF) canisters for the purpose of drop testing. The first two canisters represented a modified 24-inch diameter standardized DOE SNF canister and the second two canisters represented the Hanford Multi-Canister Overpack (MCO). The modified canisters and internals were constructed and assembled at the INEEL. The MCO internal weights were fabricated at the INEEL and assembled into two MCOs at Hanford and later shipped to the INEEL for drop test preparation. Drop testing of these four canisters was completed in August 2004 at Sandia National Laboratories. The modified canisters were dropped from 30 feet onto a flat, essentially unyielding surface, with the canisters oriented at 45 degrees and 70 degrees off-vertical at impact. One representative MCO was dropped from 23 feet onto the same flat surface, oriented vertically at impact. The second representative MCO was dropped onto the flat surface from 2 feet oriented at 60 degrees off-vertical. These drop heights and orientations were chosen to meet or exceed the Yucca Mountain repository drop criteria. This paper discusses the comparison of deformations between the actual dropped canisters and those predicted by pre-drop and limited post-drop finite element evaluations performed using ABAQUS/Explicit. Post-drop containment of all four canisters, demonstrated by way of helium leak testing, is also discussed.
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