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Tube Burst Response Of Irradiated Zircaloy Spent Fuel Cladding
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Book Synopsis Tube-Burst Response of Irradiated Zircaloy Spent-Fuel Cladding by : AA. Bauer
Download or read book Tube-Burst Response of Irradiated Zircaloy Spent-Fuel Cladding written by AA. Bauer and published by . This book was released on 1979 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: Transient-heating tube-burst tests were conducted on a 50.8-cm length of Zircaloy cladding obtained from spent-fuel rods irradiated in the H. B. Robinson power reactor to a peak burn-up of 30 MWD/kg. Internal electrical resistance heaters were used to achieve nominal heating rates of 28°C/s for the tests. The tests were conducted in steam and the independent experimental variable was the initial level of helium pressurization.
Book Synopsis Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 by :
Download or read book Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 written by and published by . This book was released on 1978 with total page 40 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 by : Arthur A. Bauer
Download or read book Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 written by Arthur A. Bauer and published by . This book was released on 1978 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Zirconium in the Nuclear Industry by : J. H. Schemel
Download or read book Zirconium in the Nuclear Industry written by J. H. Schemel and published by ASTM International. This book was released on 1979 with total page 656 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Zirconium in the Nuclear Industry by :
Download or read book Zirconium in the Nuclear Industry written by and published by ASTM International. This book was released on with total page 627 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 by : Larry M. Lowry
Download or read book Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 written by Larry M. Lowry and published by . This book was released on 1980 with total page 68 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Evaluation of Properties of Irradiated Zircaloy-2 Pressure Tube from KER Loop 1 by : L. J. Defferding
Download or read book Evaluation of Properties of Irradiated Zircaloy-2 Pressure Tube from KER Loop 1 written by L. J. Defferding and published by . This book was released on 1962 with total page 56 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 by : Larry M. Lowry
Download or read book Evaluating Strength and Ductility of Irradiated Zircaloy, Task 5 written by Larry M. Lowry and published by . This book was released on 1980 with total page 52 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Deformation and Fracture Characteristics of Spent Zircaloy Fuel Cladding by :
Download or read book Deformation and Fracture Characteristics of Spent Zircaloy Fuel Cladding written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: For a better understanding of Zircaloy fuel-rod failure by the pellet-cladding interaction (PCI) phenomenon, a mechanistic study of deformation and fracture behavior of spent power reactor fuel cladding under simulated PCI conditions was conducted. Zircaloy-2 cladding specimens, obtained from fuel assemblies of operating power reactors, were deformed to fracture at 325°C by internal gas pressurization in the absence of fission product simulants. Fracture characteristics and microstructures were examined via SEM, TEM, and HVEM. Numerous dislocation tangles and cell structures, observed in TEM specimens of cladding tubes that failed in a ductile manner, were consistent with SEM observations of a limited number of dimples characteristic of microvoid coalescence. A number of brittle-type failures were produced without the influence of fission product simulants. The brittle cracks occurred near the areas compressed by the Swagelok fittings of the internally pressurized tube and propagated from the outer to the inner surface. Since the outer surface was isolated and maintained under a flowing stream of pure helium, it is unlikely that the brittle-type failure was influenced by any fission product traces. SEM fractography of the brittle-type failure revealed a large area of transgranular pseudocleavage with limited areas of ductile fluting, which were similar in appearance to the surfaces produced by in-reactor PCI-type failures. A TEM evaluation of the cladding in the vicinity of the through-wall crack revealed numerous locations that contained an extensive amount of second-phase precipitate (Zr3O). We believe that the brittle-type failures of the irradiated spent fuel cladding in the stress rupture experiments are associated with segregation of oxygen, which leads to the formation of the order structure, an immobilization of dislocations, and minimal plastic deformation in the material.
Book Synopsis Burst Testing of Irradiated Zircaloy Tubing by : V. E. Kahle
Download or read book Burst Testing of Irradiated Zircaloy Tubing written by V. E. Kahle and published by . This book was released on 1960 with total page 32 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Burst Testing of Zircaloy Cladding from Irradiated Pickering-Type Fuel Bundles by : DG. Hardy
Download or read book Burst Testing of Zircaloy Cladding from Irradiated Pickering-Type Fuel Bundles written by DG. Hardy and published by . This book was released on 1973 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: Closed-end burst tests were conducted on sections of fuel cladding cut from five Pickering-type fuel bundles after exposures of 2.5 to 4.8 x 1024 n/m2 (>1 MeV) at 300 C (572 F). Five batches or heat treatments of Zircaloy-2 and -4 were tested at 22 and 300 C (72 and 572 F). Two of the bundles contained an unfueled element in the center ring. The burst test results obtained from fueled and unfueled sheaths are compared with the same cladding batches irradiated to similar fluences without fuel in a fast neutron facility at 125 to 250 C (257 and 482 F) and to unirradiated controls.
Download or read book Metals Abstracts written by and published by . This book was released on 1985 with total page 1156 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis A Comparison of Transient-Heating Burst Test Results of Unirradiated and Irradiated Zircaloy-4 Fuel Rod Cladding by : LM. Lowry
Download or read book A Comparison of Transient-Heating Burst Test Results of Unirradiated and Irradiated Zircaloy-4 Fuel Rod Cladding written by LM. Lowry and published by . This book was released on 1982 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Nuclear Regulatory Commission supported an experimental program to study the mechanical properties of both unirradiated and irradiated fuel rod cladding. The program was designed to produce a mechanical property data base for use in developing modeling codes which could then be used to predict the performance of Zircaloy-4 clad fuel rods under various reactor loss-of-coolant accident (LOCA) conditions. Transient-heating burst tests were conducted at Argonne National Laboratory, Battelle Columbus Laboratories, and Oak Ridge National Laboratory as part of that program. A brief description of the testing methods, specimens, equipment, and procedures illustrates the similarities and differences in tests conducted at each laboratory. The data obtained from tests conducted in steam and utilizing test specimens with internal heaters or specimens containing alumina mandrels or pellets were compared for heating rates of 5, 28, and 55 C/sec. Both burst pressure and circumferential failure strain data versus burst temperature were plotted and compared in the range of 650 to 1250 C. It was found that testing methods, specimen oxidation layer, and irradiation effected the transient-heating burst test data.
Book Synopsis NUREG/CR. by : U.S. Nuclear Regulatory Commission
Download or read book NUREG/CR. written by U.S. Nuclear Regulatory Commission and published by . This book was released on 1981 with total page 448 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Fracture Behavior and Microstructural Characteristics of Irradiated Zircaloy Cladding by : HM. Chung
Download or read book Fracture Behavior and Microstructural Characteristics of Irradiated Zircaloy Cladding written by HM. Chung and published by . This book was released on 1987 with total page 26 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zircaloy cladding tube specimens from commercial power reactor fuel assemblies (burnup >22 MWd/kgU) have been deformed to fracture at 325°C by either the internal gas-pressurization or the expanding-mandrel technique in a helium or argon environment containing no fission product species (e.g., I, Cs, or Cd). The fracture surfaces of ten irradiated specimens fractured by internal gas pressurization were examined by scanning electron microscopy; six specimens were found to contain various degrees of the pseudocleavage feature that is characteristic of pellet-cladding interaction failures. Out of ten test specimens fractured by expanding-mandrel loading, five were found to contain regions of pseudocleavage on the fracture surfaces. The specimens exhibited "X-marks" on the outer surface and brittle incipient cracks distributed on the inner surface, which are also characteristic of pellet-cladding interaction failures. Transmission/high-voltage electron microscope examinations of the thin-foil specimens obtained from regions adjacent to the failure sites showed that the ductile-failure specimens were characterized by a high density of dislocations which showed normal ->?-type Burgers vectors. In contrast, the brittle-type specimens were characterized by comparatively few dislocations which formed substructures. The dislocations in the brittle specimens were decorated by Zr3O precipitates 2 to 6 nm in size, and cubic-ZrO2 precipitates ? 10 nm in size were also observed in high density. These observations indicate a general immobilization of the dislocations. The low-ductility brittle-type failures appear to have been produced primarily as a result of the Zr3O and cubic-ZrO2 precipitates, augmented by precipitates of bulk hydride 35 to 100 nm in size. The bulk nature of these precipitates, which contrasts with the surficial nature of monoclinic ZrO2 and ?-hydride precipitates, was indicated from stereomicroscopy of weak-beam dark-field images. In situ irradiation of the spent-fuel cladding specimens by 1-MeV electrons at 325°C indicates that the Zr3O and cubic-ZrO2 precipitation is irradiation-induced, whereas the bulk hydride is not.
Book Synopsis Hydraulic Burst Tests at Elevated Temperatures on Zircaloy Cladding from Fuel Rods Irradiated in the Winfrith SGHWR by : A. Garlick
Download or read book Hydraulic Burst Tests at Elevated Temperatures on Zircaloy Cladding from Fuel Rods Irradiated in the Winfrith SGHWR written by A. Garlick and published by . This book was released on 1980 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes by : S. Lansiart
Download or read book Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes written by S. Lansiart and published by . This book was released on 1994 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: Since nuclear electricity has a predominant share in French generating capacity, pressurized water reactors (PWRs) are required to fit grid load following and frequency control operating conditions. Consequently, cyclic stresses appear in the fuel element cladding. In order to characterize the possible resulting clad damage, fatigue tests were performed at 350°C on unirradiated material or irradiated stress relieved Zircaloy-4 tube portions, using a special device for tube fatigue by repeated pressurization.
