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Summary Report for Contract 330088-A-T9: Procurement of Zircaloy-4 Products from Superior Tube Company

Description: This report documents the initial procurement activities from the Superior Tube Company associated with production quantities of Zircaloy 4 components (Contract 330088-A-T9) and the demonstration of their capability to produce full-length getter components (Contract 406957-A-T5).
Date: June 27, 2007
Creator: Graves, Roger E.
Partner: UNT Libraries Government Documents Department

On the embrittlement of Zircaloy-4 under RIA-relevant conditions.

Description: The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding. Utilizing plane-strain ring-stretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300 C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness >100 {micro}m ({approx}700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness <90 {micro}m ({approx}600 wppm) remain ductile. The mechanism of failure is identified as strain-induced crack initiation within the hydride rim and failure within the uncracked ligament due to either a shear instability or damage-induced fracture. We also report some preliminary results of the uniaxial tensile behavior of low-Sn Zircaloy-4 cladding tubes in a cold-worked, stress-relieved condition in the transverse (hoop) direction at strain rates of 0.001/s and 0.2/s and temperatures of 26-400 C.
Date: December 19, 2001
Creator: Daum, R.S.; Majumdar, S.; Billone, M.C.; Bates, D.W.; Koss, D.A. & Motta, A.T.
Partner: UNT Libraries Government Documents Department


Description: For this study, hydrogen getter materials (Zircaloy-4 and pure zirconium) that have a high affinity for hydrogen (and low overpressure) have been investigated to determine the hydrogen equilibrium pressure on Zircaloy-4 and pure zirconium. These materials, as with most getter materials, offered significant challenges to overcome given the low hydrogen equilibrium pressure for the temperature range of interest. Hydrogen-zirconium data exists for pure zirconium at 500 C and the corresponding hydrogen overpressure is roughly 0.01 torr. This manuscript presents the results of the equilibrium pressures for the absorption and desorption of hydrogen on zirconium materials at temperatures ranging from 400 C to 600 C. The equilibrium pressures in this temperature region range from 150 mtorr at 600 C to less than 0.1 mtorr at 400 C. It has been shown that the Zircaloy-4 and zirconium samples are extremely prone to surface oxidation prior to and during heating. This oxidation precludes the hydrogen uptake, and therefore samples must be heated under a minimum vacuum of 5 x 10{sup -6} torr. In addition, the Zircaloy-4 samples should be heated at a sufficiently low rate to maintain the system pressure below 0.5 mtorr since an increase in pressure above 0.5 mtorr could possibly hinder the H{sub 2} absorption kinetics due to surface contamination. The results of this study and the details of the testing protocol will be discussed.
Date: April 3, 2012
Creator: Morgan, G. & Korinko, P.
Partner: UNT Libraries Government Documents Department


Description: Thermogravimetric analysis (TGA) and coupled Mass Spectroscopy (MS) were evaluated to determine their suitability as a quality assurance tool for surface modified nickel plated zircaloy-4 liner tubes. Samples with 0, 0.1, 0.2, 0.3, and 0.4 mils of heat treated nickel plate were tested at 330, 370, and 400 C. Not all of the samples exhibited the expected typical parabolic shaped oxidation curve. The measured weight change was consistent for the as received and 0.2 mil and the 0.4 mil surface modified samples. None of the samples were tested under aggressive enough conditions to consume the surface modified materials during the test duration. Use of the Mass Spectrometer in conjunction with the TGA did not produce valuable data and was only used for the 400 C test series; however, the TGA was valuable. The 0.1 and 0.3 mil surface modified Zr-4 samples exhibited thru surface modified layer cracks which could account for the variation in oxidation behavior. TGA tests for periods up to six hours appear viable as a method to ascertain oxidation behavior for consistent results. Additional testing of samples with known variations in surface modified layer thickness and quality is recommended as part of the QA acceptance testing.
Date: September 21, 2009
Creator: Korinko, P. & Imrich, K.
Partner: UNT Libraries Government Documents Department

Intergrannular strain evolution in a zircaloy-4 alloy with Widmanstatten microstructure

Description: A Zircaloy-4 alloy with Widmanstatten-Basketweave microstructure and random texture has been used to study the deformation systems responsible for the polycrystalline plasticity at the grain level. The evolution of internal strain and bulk texture is investigated using neutron diffraction and an elasto-plastic self-consistent (EPSC) modeling scheme. The macroscopic stress-strain behavior and intergranular (hkil-specific) strain development, parallel and perpendicular to the loading direction, were measured in-situ during uniaxial tensile loading. Then, the EPSC model was employed to simulate the experimental results. This modeling scheme accounts for the thermal anisotropy; elastic-plastic properties of the constituent grains; and activation, reorientation, and stress relaxation associated with twinning. The agreement between the experiment and the model will be discussed as well as the critical resolved shear stresses (CRSS) and the hardening coefficients obtained from the model.
Date: January 1, 2009
Creator: Clausen, Bjorn; Vogel, Sven C; Garlea, Eena; Choo, Hahn; Pang, Judy W L & Kenik, Edward A
Partner: UNT Libraries Government Documents Department

The long range migration of hydrogen through Zircaloy in response to tensile and compressive stress gradients

Description: Zircaloy-4, which is used widely as a core structural material in pressurized water reactors (PWRs), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and zirconium hydride phases precipitate after the Zircaloy-4 lattice becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4, degrading its mechanical performance as a structural material. Because hydrogen can move rapidly through the Zircaloy-4 lattice, the potential exists for large concentrations of hydride to accumulate in local regions of a Zircaloy component remote from its point of entry into the component. Much has been reported in the literature regarding the long range migration of hydrogen through Zircaloy under concentration gradients and temperature gradients. Relatively little has been reported, however, regarding the long range migration of hydrogen under stress gradients. This paper presents experimental results regarding the long range migration of hydrogen through Zircaloy in response to both tensile and compressive stress gradients. The importance of this driving force for hydrogen migration relative to concentration and thermal gradients is discussed.
Date: November 1, 1998
Creator: Kammenzind, B.F.; Berquist, B.M.; Bajaj, R.; Kreyns, P.H. & Franklin, D.G.
Partner: UNT Libraries Government Documents Department

Spent fuel cladding corrosion under tuff repository conditions: initial observations

Description: The Westinghouse Hanford Company program is investigating corrosion and stress corrosion cracking of Zircaloy-2 and 4 in two model tuff repository environments using an experimental approach in which the repository environment is reproduced as accurately as possible, including temperature, radiation field, water chemistry and materials associations. Post-experimental sample evaluation utilizes or will utilize sophisticated SEM/STEM, Auger surface analysis/ion milling, and trace element release to detect, locate and measure the effects of corrosion. The experiments themselves are being conducted using actual spent fuel and repository materials at repository conditions. The short experimental time (i.e., one year) is being compensated for by sensitive measuring techniques. Characterization of any corrosion found will be used to understand the mechanisms involved for extrapolation purposes. The initial evaluation of samples from two, six, and 12-month electrochemical corrosion experiments indicated no Zircaloy-4 corrosion at a detection sensitivity of 1 to 2 {mu}m of corrosion per year. To improve the sensitivity of the experiment, baseline conditions (e.g., beginning with a polished metal surface) will need to be established that are expected to make it possible to resolve corrosion on the scale of hundreds of angstroms. Examples are the development of such measurements as film depth determination via Auger surface analysis/ion milling and Zr and {sup 14}C released into the aqueous corrosion environment. Characterization of any corrosion found will be used to understand the mechanisms involved. This will allow extrapolation of results to predict cladding lifetime under repository conditions. 3 refs.
Date: June 1, 1985
Creator: Smith, H.D. & Oversby, V.M.
Partner: UNT Libraries Government Documents Department

Amorphization of Laves-Phase Precipitates in Zircaloy-4 by Neutron Irradiation

Description: Examination of corrosion coupons by transmission electron microscopy after their exposure in the Idaho Advanced Test Reactor (ATR) has broadened the Zircaloy-4 precipitate-amorphization database and validated a new kinetic model for previously unavailable values of temperature and fast-neutron flux. The model describes the amorphization of Zr(Fe,Cr){sub 2} intermetallic precipitates in zirconium alloys as a dynamic competition between radiation damage and thermal annealing that leaves some iron atoms available for flux-assisted diffusion to the zirconium matrix. It predicts the width of the amorphous zone as a function of neutron flux (E>1 MeV), temperature, and time. In its simplest form, the model treats the crystalline/amorphous and precipitate/matrix interfaces as parallel planes, and its accuracy decreases for small precipitates and high fluence as the amorphous-zone width approaches precipitate dimensions. The simplest form of the model also considers diffusion to be rate-determining. This is an accurate approximation for steady-state conditions or slow changes in flux and temperature, but inappropriate for the analysis of faster transients. The paper addresses several difficulties inherent in measuring amorphous-zone width, and utilizes the expanded database to evaluate the improvements in predictive accuracy available through both conversion of the model to spherical coordinates and extension of its time dependency.
Date: April 23, 1999
Creator: Peters, H.R.; Taylor, D.F. & Yang, Walter J.S.
Partner: UNT Libraries Government Documents Department

Mapping Flow Localization Processes in Deformation of Irradiated Reactor Structural Alloys

Description: Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the metal by causing premature plastic instability failure or by inducing the formation of cracks. Irradiation with neutrons hardens a metal and makes it more prone to deformation by strain localization. Although this has been known since the earliest days of radiation damage studies, a full measure of the connection between neutron irradiation hardening and strain localization is wanting, particularly in commercial alloys used in the construction of nuclear reactors. Therefore, the goal of this project is to systematically map the extent of involvement of strain localization processes in plastic deformation of three reactor alloys that have been neutron irradiated. The deformation processes are to be identified and related to changes in the tensile properties of the alloys as functions of neutron fluence (dose) and degree of plastic strain. The intent is to define the role of strain localization in radiation embrittlement phenomena. The three test materials are a tempered bainitic A533B steel, representing reactor pressure vessel steel, an annealed 316 stainless steel and annealed Zircaloy-4 representing reactor internal components.
Date: July 18, 2002
Creator: Farrell, K.
Partner: UNT Libraries Government Documents Department

Gallium interactions with Zircaloy

Description: This study focuses on the effects of gallium ion implantation into zircaloy cladding material to investigate the effects that gallium may have in a reactor. High fluence ion implantation of Ga ions was conducted on heated Zircaloy-4 in the range of 10{sup 16}--10{sup 18} Ga ions/cm2. Surface effects were studied using SEM and electron microprobe analysis. The depth profile of Ga in the Zircaloy was characterized with Rutherford backscattering and SIMS techniques. Results indicate that the Zirc-4 is little affected up to a fluence of 10{sup 17} Ga ions/cm{sup 2}. After implantation of 10{sup 18} Ga ions/cm{sup 2}, sub-grain features on the order of 2 {micro}m were observed which may be due to intermetallic compound formation between Ga and Zr. For the highest fluence implant, Ga content in the Zirc-4 reached a saturation value of between 30 and 40 atomic %; significant enhanced diffusion was observed but gallium was not seen to concentrate at grain boundaries.
Date: January 1, 1999
Creator: Woods, A. L. & West, M. K.
Partner: UNT Libraries Government Documents Department

Gallium interactions with zircaloy cladding

Description: The effects of Ga from weapons-grade plutonium MOX fuel on zircaloy-IV cladding during power reactor operation have been simulated by implantations of 100 keV Ga-69 ions into a polished zircaloy-IV sample while the sample was maintained at a typical cladding temperature of 375{degrees}C. Analyses were based on scanning electron microscopy, Rutherford backscattering of 280 keV He-3 ions, and secondary ion mass spectroscopy. Subgrains at the zircaloy-IV surface formed at a Ga fluence equivalent to total release of approximately 12 ppm by weight of Ga from the fuel. The subgrains may be an intermetallic compound of Zr{sub 2}Ga. Enhanced diffusion of Ga was observed, but Ga concentrations decreased 3 orders of magnitude over a depth of 3000 {angstrom}.
Date: May 1, 1998
Creator: Hart, R. R.; Rennie, J.; Aucoin, K. & West, M.
Partner: UNT Libraries Government Documents Department

Cladding embrittlement during postulated loss-of-coolant accidents.

Description: The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.
Date: July 31, 2008
Creator: Billone, M.; Yan, Y.; Burtseva, T. & Daum, R.
Partner: UNT Libraries Government Documents Department