Download or read book Characteristics of Autoclave and In reactor Nodular Corrosion of Zircaloys written by Y. H. Jeong and published by . This book was released on 1990 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Characteristics of Autoclave and In reactor Nodular Corrosion of Zircaloys written by and published by . This book was released on 1990 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nodular corrosion characteristics of recrystallized Zircaloy-4 were investigated in static autoclave tests at 500°C and 10.3 MPa. The roles of annealing temperature, cooling rate after beta-treating at 1050°C, cold work, and surface treatment in corrosion tests were correlated with the results of microstructural characterization by scanning and transmission electron microscopies. A good correlation was obtained between average size of intermetallic precipitates and weight gain, in contrast to nodule coverage and nodule number density. These results could be best explained by the hypothesis that nodules nucleate in local regions that are depleted of Fe and Cr alloying elements. Some observations were inconsistent with the premise that nodules nucleate on or near intermetallic precipitates. Nodular corrosion characteristics and microstructures of commercial Zircaloy-2 cladding of fuel and gadolinia rods, obtained from several BWRs after burnup of 11--30 MWd/kgU, were also examined. Partial amorphization of intermetallic precipitates in BWR Zircaloy-2, and virtual dissolution and in an extreme case spinodal- like fluctuations of dissolved alloying elements in PWR Zircaloy-4 cladding were observed. Occurrence of nodular oxidation of Zircaloy-2 in BWRs could best be correlated to average size of intermetallic precipitates before irradiation and to fuel cladding operating temperature. For an intermetallic size range of 250--700 nm, nodular oxides were observed at 288°C, but only thick uniform oxide was observed at 307°C. 53 refs., 14 figs., 1 tab.
Download or read book Zirconium in the Nuclear Industry written by Craig M. Eucken and published by ASTM International. This book was released on 1991 with total page 794 pages. Available in PDF, EPUB and Kindle. Book excerpt: The proceedings of the Ninth International Symposium on [title], held in Kobe, Japan, November 1990, address current trends in the development, performance, and fabrication of zirconium alloys for nuclear power reactors. the bulk of the most recent work on zirconium alloy behavior has concerned corr
Download or read book Nodular Corrosion of the Zircaloys written by and published by . This book was released on 1976 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Oxide nodules form on Zircaloy nuclear components under irradiation. Similar nodules were observed on Zircaloy coupons in cold rolled or extended conditions after autoclave treatments at 475 and 500/sup 0/C in steam at 1500 to 1700 psi. Stages of nodular corrosion in the autoclave were nodule nucleation, growth, coalescence, propagation to accelerated uniform corrosion, and complete specimen oxidation. Observations on BWR fuel rods suggest that a similar progressive attack has occurred; however, in no case has the in-reactor attack progressed to the stage of complete fuel rod failure. Recent autoclave tests confirmed the nodular character of the attack on cold-worked materials. Alpha anneals (up to 790/sup 0/C) did not consistently suppress the nodular attack. However, alpha + beta (840/sup 0/C) and beta (1010 and 1040/sup 0/C) anneals did suppress the attack if followed by a fast cool. The efficacy of the anneals applied similarly to Zircaloy-2 and Zircaloy-4. Stresses associated with U-bend specimens and heavy (86 percent) cold work did not enhance the nodular attack before stress relief occurred. The nodular attack on reactor components appears to depend on nuclear flux, and develops in oxygenated reactor coolants, principally in the vicinity of fuel rod spacers. Experience with irradiated specimens in reactor loops suggests that uniform concentrations of dissolved oxygen alone do not initiate nodular attack. Localized water chemistry associated with flow disturbances or, in some cases, dissimilar metals in fuel spacers, may be factors in the nodular attack in-reactor.
Download or read book Corrosion Testing of Zircaloy 2 and Zircaloy 3 written by Stanley Kass and published by . This book was released on 1957 with total page 64 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book In pile autoclave Test of Zircaloy 2 Corrosion in UO2 NO3 2 Solution and of Solution Stability written by R. J. Davis and published by . This book was released on 1962 with total page 32 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nodular Corrosion of the Zircaloys written by RM. Horton and published by . This book was released on 1977 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Oxide nodules form on Zircaloy nuclear components under irradiation. Similar nodules were observed on Zircaloy coupons in cold-rolled or extruded conditions after autoclave treatments at 475 and 500 °C in steam at 1500 to 1700 psi. The stages of nodular corrosion in the autoclave were: nodule nucleation, growth, coalescence, propagation to accelerated uniform corrosion, and complete specimen oxidation. Observations on boiling water reactor (BWR) fuel rods suggest that a similar progressive attack has occurred; however, in no case has the in-reactor attack appeared to progress to the stage of complete component failure.
Download or read book Scientific and Technical Aerospace Reports written by and published by . This book was released on 1991 with total page 280 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Zirconium in the Nuclear Industry written by Gerry D. Moan and published by ASTM International. This book was released on 2002 with total page 891 pages. Available in PDF, EPUB and Kindle. Book excerpt: Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.
Download or read book Parametric Study of an Autoclave Test for Nodular Corrosion written by JH. Schemel and published by . This book was released on 1987 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: The use of a steam corrosion autoclave test as a quality control tool for Zircaloy products has been practiced for many years. The most common form of the test is exposure to steam at 672 K, 10.3 MPa (750°F, 1500 psi) for three days. Some years ago it was found that this test did not predict susceptibility to the occurrence nodular corrosion in boiling water reactors (BWR). A number of investigators found that a test at 723 to 773 K would produce nodular corrosion on Zircaloy products and therefore used the test as an index for research work on Zircaloy materials. More recently the test has been proposed as a quality control test on each lot of material. This study of test parameters and interlaboratory variability was undertaken as a necessary part of establishing an ASTM standard for such a test. Temperature of the test was found to be the most significant variable studied, and interlaboratory variation was relatively high when compared with the common 672 K test.
Download or read book Zirconium in the Nuclear Industry Proceedings of the Third International Conference written by and published by ASTM International. This book was released on with total page 681 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book A Systematic Survey of the Factors Affecting Zircaloy Nodular Corrosion written by K. Abe and published by . This book was released on 1989 with total page 24 pages. Available in PDF, EPUB and Kindle. Book excerpt: Out-of-reactor corrosion tests in 10.3-MPa steam at temperatures between 490 and 530°C have been carried out on Zircaloy-2 tube specimens of various fabrication histories including archive specimens of the same tube lots as were used in the commercial boiling water reactor (BWR) to get data for comparison between in- and out-of-reactor nodular corrosion performances. Test conditions, such as temperature and duration, as well as material parameters, such as chemical composition, texture, and precipitates, were examined to clarify the relation to the nodular corrosion susceptibility.
Download or read book Nodular Corrosion of Zircaloy 2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments written by A. Garlick and published by . This book was released on 1977 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: In steam generating heavy water reactors (SGHWR), nodular or "patch-type" corrosion is the dominant form of attack on Zircaloy, as it is in other boiling or oxygenated reactor environments. The most severe oxidation occurs close to the stainless steel spacer grids, although nodules are seen elsewhere on cladding and on non-fueled components, a maximum thickness of 160 ?m at 20 MWd/kg uranium local burn-up having been recorded to date. The extensive data from post-irradiation examination of SGHWR fuel elements are reviewed and discussed from the points of view of determination of rate laws for oxide growth and the establishment of corrosion mechanisms.
Download or read book Correlation of Autoclave Testing of Zircaloy 4 to In Reactor Corrosion Performance written by RA. Parkins and published by . This book was released on 1994 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: As the service environment for core materials for Pressurized Water Reactors (PWRs) becomes more severe due to increases in water temperature, changes in water chemistry, longer cycle lengths and higher burnups; the corrosion performance of Zircaloy-4 which is used for fuel cladding and structural components becomes more important. The standard three day, 400°C steam autoclave test which has been used to evaluate the uniform corrosion resistance of Zircaloy-4 is no longer sufficient. Some cladding which has given acceptable results in this test has not exhibited satisfactory in-reactor corrosion performance. A series of cladding types for which in-reactor corrosion measurements are available has been subjected to extended autoclave testing in steam at 400 and 415°C. A better correlation to the in-reactor performance is obtained if the 400 °C test is extended to at least 100 days or if the test temperature is raised to 415°C. At 415°C, a test duration of three days is best for evaluating Zircaloy-4 since extended time does not improve the correlation.
Download or read book Zirconium in the Nuclear Industry written by R. B. Adamson and published by ASTM International. This book was released on 1987 with total page 832 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Influence of Chemical Composition and Manufacturing Variables on Autoclave Corrosion of the Zircaloys written by PT. Finden and published by . This book was released on 1989 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: Previous investigators have reported the separate influences of chemical composition and heat treatment on the nodular and uniform corrosion of zirconium-based alloys. However, many of these compositions lie outside the allowable Zircaloy ranges required by most nuclear fuel specifications. The current study shows that the corrosion performance of the Zircaloys can be optimized by adjusting chemical compositions within current ASTM ranges, and examines the influence of composition on response to heat treatment. It is believed that this approach will yield improved corrosion performance and results in a product that can be introduced for immediate nuclear service application.
Download or read book Further Evidence of Zircaloy Corrosion in Fuel Elements Irradiated in a Steam Generating Heavy Water Reactor written by A. Garlick and published by . This book was released on 1979 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: An earlier paper (Trowse et al, ASTM STP 633) described the nodular corrosion of Zircaloy and other zirconium alloys in a steam generating heavy water reactor (SGHWR) and related environments. The observations have now been extended to more highly irradiated elements up to the design burn-up of 20 MWD/kgU mean. Maximum observed local oxide thicknesses were less than the earlier predictions of 200 ?m for SGHWR cladding at the design burn-up and are acceptable from fuel performance considerations.
