30 research outputs found
Ripening-induced changes in microbial groups of artisanal Sicilian goats' milk cheese.
Changes in the microbial flora of "Caprino dei Nebrodi", a raw goat's milk cheese produced in Sicily, were studied during ripening. From 2 batches of cheese, 4 samples were taken at day 0, 2, 15, and 30 of ripening. Also, samples of curd and milk used in the manufacturing process were analyzed. By the end of the ripening process (day 30), high log10 cfu/g were found for Lactobacilli (7.20), Lattococci (7.10), and Enterococci (7.00), whereas counts of Enterobacteriaceae (3.91), Escherichia coli (3.30), and Staphylococcus (3.89) were found to be lower. The study provides useful information on the microbiological properties of "Caprino dei Nebrodi" cheese, and the results obtained suggest that in order to increase the quality of this artisanal product, it is necessary to improve the sanitary conditions of milking and cheese-making. The study was intended as a preliminary step towards the isolation and identification of bacterial species found in this type of goat's cheese
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INTOR first wall/blanket/shield activity
The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory
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TRIO-01 experiment: in-situ tritium recovery results
The TRIO-01 experiment was designed to test in-situ tritium recovery and heat transfer performance of a candidate solid breeder, ..gamma..-LiAlO/sub 2/. The results showed that nearly all the tritium generated was recovered. Only < 0.1 wppM tritium remained in the solid after irradiation testing. The heat transfer performance showed that temperature profiles can be effectively controlled
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Irradiation Performance of U-Pu-Zr Metal Fuels for Liquid-Metal-Cooled Reactors
This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result`s from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of {approx}75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of {approx}1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of {approx}11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction
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Steady-state deformation of some lithium ceramics
The stress-strain behavior of Li/sub 2/O, LiAlO/sub 2/ and Li/sub 2/ZrO/sub 3/ polycrystals, with densities varying from 0.70 to 0.95 of the theoretical, has been measured in constant-crosshead-speed compression tests at temperatures of 700 to 1000/sup 0/C with strain rates ranging from about 10/sup -6/ to 10/sup -4/ s/sup -1/. A steady-state stress, sigma/sub s/, for which the work-hardening rate becomes zero, was achieved. These results, therefore, yield information equivalent to that obtained from creep experiments. Limited data on LiAlO/sub 2/ and Li/sub 2/ZrO/sub 3/ were obtained. Nevertheless, under comparable conditions the lithium aluminate and zirconate were considerably stronger than the Li/sub 2/O. This finding may be related to differences in crystal structure. It is, however, likely that in operation as a function breeder blanket material, the oxide will swell whereas the aluminate and the zirconate will crack. 5 refs., 6 figs., 1 tab
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Examination of spent PWR fuel rods after 15 years in dry storage.
Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas release from the fuel pellets occurred during the thermal benchmark tests or storage. Measurements of the cladding outer-diameter, oxide thickness and wall thickness are in the expected range for cladding of the Surry exposure. The measured hydrogen content is consistent with the oxide thickness. The volume of hydrides varies azimuthally around the cladding, but there is little variation across the thickness, of the cladding. It is most significant that all of the hydrides appear to have retained the circumferential orientation typical of prestorage PWR fuel rods
Laparoscopic-assisted Retropubic Midurethral Sling Placement: A Technique to Avoid Major Complications
Study Objective To describe a technique for the safe placement of retropubic midurethral slings in patients undergoing concomitant laparoscopic surgery in order to avoid major complications associated with this procedure such as bladder perforation and retropubic hematomas. Design Step-by-step video demonstration of the technique. Setting A university tertiary care hospital. Patients Patients with an indication for retropubic midurethral sling placement because of recurrent stress urinary incontinence, intrinsic sphincter deficiency, or severe pelvic organ prolapse in whom a concomitant laparoscopic surgery has to be performed for other medical conditions. Intervention Laparoscopic opening and dissection of the Retzius space and insertion of the sling under a laparoscopic view of this space. Measurements and Main Results This technique has been mainly used in patients undergoing laparoscopic pelvic organ prolapse repair. No complications have been identified so far, even in high-risk patients such as those with previous Burch colposuspension. Conclusion This is a simple and reproducible technique for preventing major complications associated with retropubic sling placement in patients undergoing laparoscopic surgery for other reasons. It also permits the immediate detection and even resolution of complications in case any arise. Even high-risk patients may be safely approached
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Modified ring stretch tensile testing of Zr-1Nb cladding.
In a round robin effort between the US Nuclear Regulatory Commission, Institut de Protection et de Surete Nucleaire in France, and the Russian Research Centre-Kurchatov Institute, Argonne National Laboratory conducted 16 modified ring stretch tensile tests on unirradiated samples of Zr-1Nb cladding, which is used in Russian VVER reactors. Tests were conducted at two temperatures (25 and 400 C) and two strain rates (0.001 and 1 s{sup {minus}1}). At 25 C and 0.001 s{sup {minus}1}, the yield strength (YS), ultimate tensile strength (UTS), uniform elongation (UE), and total elongation (TE) were 201 MPa, 331 MPa, 18.2%, and 57.6%, respectively. At 400 C and 0.001 s{sup {minus}1}, the YS, UTS, UE, and TE were 109 MPa, 185 MPa, 15.4%, and 67.7%, respectively. Finally, at 400 C and 1 s{sup {minus}1}, the YS, UTS, UE, and TE were 134 MPa, 189 MPa, 18.9%, and 53.4%, respectively. The high strain rate tests at room temperature were not successful. Test results proved to be very sensitive to the amount of lubrication used on the inserts; because of the large contact area between the inserts and specimen, too little lubrication leads to significantly higher strengths and lower elongations being reported. It is also important to note that only 70 to 80% of the elongation takes place in the gauge section, depending on specimen geomeq. The appropriate percentage can be estimated from a simple model or can be calculated from finite-element analysis
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TRIO experiment
The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an anaytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion