67 Matching Results

Search Results

Advanced search parameters have been applied.

Phase transformations in neutron-irradiated Zircaloys

Description: Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after approx.3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr/sub 3/O and cubic-ZrO/sub 2/ particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/)/sub 2/ and Zr/sub 2/(Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of approx.4 x 10/sup 21/ ncm/sup -2/ in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs.
Date: April 1, 1986
Creator: Chung, H.M.
Partner: UNT Libraries Government Documents Department

Structure of high-burnup-fuel Zircaloy cladding. [PWR; BWR]

Description: Zircaloy cladding from high-burnup (> 20 MWd/kg U) fuel rods in light-water reactors is characterized by a high density of irradiation-induced defects (RID), compositional changes (e.g., oxygen and hydrogen uptake) associated with in-service corrosion, and geometrical changes produced by creepdown, bowing, and irradiation-induced growth. During a reactor power transient, the cladding is subject to localized stress imposed by thermal expansion of the cracked fuel pellets and to mechanical constraints imposed by pellet-cladding friction. As part of a program to provide a better understanding of brittle-type failure of Zircaloy fuel cladding by pellet-cladding interaction (PCI) phenomenon, the stress-rupture properties and microstructural characteristics of high-burnup spent fuel cladding have been under investigation. This paper reports the results of the microstructural examinations by optical microscopy, scanning (SEM), 100-keV transmission (TEM), and 1 MeV high-voltage (HVEM) electron microscopies of the fractured spent fuel cladding with a specific empahsis on a correlation of the structural characteristics with the fracture behavior.
Date: June 1, 1983
Creator: Chung, H.M.
Partner: UNT Libraries Government Documents Department

Thermal aging of some decommissioned reactor components and methodology for life prediction

Description: Since a realistic aging of cast stainless steel components for end-of-life or life-extension conditions cannot be produced, it is customary to simulate the thermal aging embrittlement by accelerated aging at /approximately/400/degree/C. In this investigation, field components obtained from decommissioned reactors have been examined after service up to 22 yr to provide a benchmark of the laboratory simulation. The primary and secondary aging processes were found to be identical to those of the laboratory-aged specimens, and the kinetic characteristics were also similar. The extent of the aging embrittlement processes and other key factors that are known to influence the embrittlement kinetics have been compared for the decommissioned reactor components and materials aged under accelerated conditions. On the basis of the study, a mechanistic understanding of the causes of the complex behavior in kinetics and activation energy of aging (i.e., the temperature dependence of aging embrittlement between the accelerated and reactor-operating conditions) is presented. A mechanistic correlation developed thereon is compared with a number of available empirical correlations to provide an insight for development of a better methodology of life prediction of the reactor components. 18 refs., 18 figs., 5 tabs.
Date: March 1, 1989
Creator: Chung, H.M.
Partner: UNT Libraries Government Documents Department

Assessment of void swelling in austenitic stainless steel PWR core internals.

Description: As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and hence, high ...
Date: January 31, 2006
Creator: Chung, H. M. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

Description: Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850/sup 0/C or extrapolated from the low-temperature data obtained at <1500/sup 0/C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, rather than in an unlimited flux of pure steam. New results of Zircaloy oxidation measurements in various mixtures of hydrogen and steam are reported for >1500/sup 0/C. The results show significantly smaller oxidation and, hence, hydrogen-generation rates in the mixture, compared with those obtained in pure steam. It is also shown that a significant fraction of hydrogen, generated as a result of Zircaloy oxidation, is dissolved in the cladding material itself, which prevents that portion of hydrogen from reaching the containment building space. Implications of these findings are discussed in relation to a more realistic method of quantifying the hydrogen source term for a degraded-core accident analysis.
Date: February 1, 1983
Creator: Chung, H.M. & Thomas, G.R.
Partner: UNT Libraries Government Documents Department

Hydride-related degradation of spent-fuel cladding under repository conditions

Description: This report summarizes results of an analysis of hydride-related degradation of commercial spent-nuclear-fuel cladding under repository conditions. Based on applicable laboratory data on critical stress intensity obtained under isothermal conditions, occurrence of delayed hydride cracking from the inner-diameter side of cladding is concluded to be extremely unlikely. The key process for potential initiation of delayed hydride cracking at the outer-diameter side is long-term microstructural evolution near the localized regions of concentrated hydrides, i.e., nucleation, growth, and cracking of hydride blisters. Such locally concentrated hydrides are, however, limited to some high-burnup cladding only, and the potential for crack initiation and propagation at the outer-diameter side is expected to be insignificant for most spent fuels. Some degree of hydride reorientation could occur in high-burnup spent-fuel cladding. However, even if hydride reorientation occurs, accompanying stress-rupture failure in spent-fuel cladding is unlikely to occur.
Date: April 3, 2000
Creator: Chung, H. M.
Partner: UNT Libraries Government Documents Department

Fundamental metallurgical aspects of axial splitting in zircaloy cladding

Description: Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10{sup 21} n cm{sup {minus}2} to 5.9 x 10{sup 21} n cm{sup {minus}2} (E &gt; 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed.
Date: April 3, 2000
Creator: Chung, H. M.
Partner: UNT Libraries Government Documents Department

Tensile impact properties of vanadium-base alloys irradiated at &lt;430{degree}C.

Description: Tensile and impact properties were investigated at &lt;430 C on V-Cr-Ti, V-Ti-Si, and V-Ti alloys after irradiation to {approx}2-46 dpa at 205-430 C in lithium or helium in the Fast Flux Test Facility (FFTF), High Flux Isotope Reactor (HFIR), Experimental Breeder Reactor II (EBR-II), and Advanced Test Reactor (ATR). A 500-kg heat of V-4Cr-4Ti exhibited high ductile-brittle transition temperature and minimal uniform elongation as a result of irradiation-induced loss of work-hardening capability. Work-hardening capabilities of 30- and 100-kg heats of V-4Cr-4Ti varied significantly with irradiation conditions, although the 30-kg heat exhibited excellent impact properties after irradiation at {approx}390-430 C. The origin of the significant variations in the work-hardening capability of V-4Cr-4Ti is not understood, although fabrication variables, annealing history, and contamination from the irradiation environment are believed to play important roles. A 15-kg heat of V-3Ti-1Si exhibited good work-hardening capability and excellent impact properties after irradiation at {approx}390-430 C. Helium atoms, either charged dynamically or produced via transmutation of boron in the alloys, promote work-hardening capability in V-4Cr-4Ti and V-3Ti-1Si.
Date: May 18, 1998
Creator: Chung, H. M.
Partner: UNT Libraries Government Documents Department

Mechanical properties and microstructural characteristics of laser and electron-beam welds in V-4Cr-4Ti.

Description: Mechanical properties and microstructural characteristics of laser and electron-beam welds of a 500-kg heat of V4Cr4Ti were investigated in as-welded condition and after postwelding heat treatment by impact testing, microhardness measurement, optical microscopy, X-ray diffraction, and transmission electron microscopy (TEM). Ductile-brittle transition temperatures of the laser and electron-beam welds were significantly higher than that of the base metal. However, excellent impact properties could be restored in both types of welds by postwelding annealing at 1000 C for 1 h in vacuum. Analysis by TEM revealed that annealed weld zones were characterized by extensive networks of fine V(C,O,N) precipitates, which clean away O, C, and N interstitial from the grain matrices. This process is accompanied by simultaneous annealing-out of the dense dislocations present in the weld zone. This finding could be useful in identifying an optimal welding procedure by controlling and adjusting the cooling rate of the weld zone by an innovative method to maximize the precipitation of V(C,O,N).
Date: May 18, 1998
Creator: Chung, H. M.
Partner: UNT Libraries Government Documents Department

TEM (transmission electron microscopy), APFIM (atom-probe field ion microscopy), and SANS (small-angle neutron scattering) examination of aged duplex stainless steel components from some decommissioned reactors

Description: The present investigation indicates that the primary embrittlement processes of the CF-8 grade cast stainless steel components during extended reactor service are spinodal decomposition of the ferrite phase and M/sub 23/C/sub 6/ carbide precipitation on the austenite-ferrite boundaries. The ferrite hardness measured for the Shippingport reactor valves appears to reflect the different extent of spinodal decomposition among the different valves which contain slightly different Cr contents. G-phase precipitation was minimal compared to that during accelerated aging of CF-8 steel in the laboratory (i.e., near 400/degree/C). This indicates that the activation energy may be strongly influenced by the synergism among the G-phase precipitation, carbide formation, and spinodal decomposition. 13 refs., 2 figs.
Date: December 1, 1987
Creator: Chung, H.M. & Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Kinetics and mechanism of thermal aging embrittlement of duplex stainless steels

Description: Microstructural characteristics of long-term-aged cast duplex stainless steel specimens from eight laboratory heats and an actual component from a commercial boiling water reactor have been investigated by scanning electron microscopy (SEM), transmission electron microscopy (TEM), small angle neutron scattering (SANS), and atom probe field ion microscopy (APFIM) techniques. Three precipitate phases, i.e., Cr-rich ..cap alpha..' and the Ni- and Si-rich G phase, and ..gamma../sub 2/ austenite, have been identified in the ferrite matrix of the aged specimens. For CF-8 grade materials, M/sub 23/C/sub 6/ carbides were identified on the austenite-ferrite boundaries as well as in the ferrite matrix for aging at greater than or equal to 450/sup 0/C. It has been shown that Si, C, and Mo contents are important factors that influence the kinetics of the G-phase precipitation. However, TEM and APFIM analyses indicate that the embrittlement for less than or equal to400/sup 0/C aging is primarily associated with Fe and Cr segregation in ferrite by spinodal decomposition. For extended aging, e.g., 6 to 8 years at 350 to 400/sup 0/C, large platelike ..cap alpha..' formed by nucleation and growth from the structure produced by the spinodal decomposition. The Cr content appears to play an important role either to promote the platelike ..cap alpha..' (high Cr content) or to suppress the ..cap alpha..' in favor of ..gamma../sub 2/ precipitation (low Cr). Approximate TTT diagrams for the spinodal, ..cap alpha..', G, ..gamma../sub 2/, and the in-ferrite M/sub 23/C/sub 6/ have been constructed for 250 to 450/sup 0/C aging. Microstructural modifications associated with a 550/sup 0/C reannealing and a subsequent toughness restoration are also discussed. It is shown that the toughness restoration is associated primarily with the dissolution of the Cr-rich region in ferrite.
Date: June 1, 1987
Creator: Chung, H.M. & Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Aging degradation of cast stainless steels: Effects on mechanical properties

Description: A program is being conducted to investigate the significance of in-service embrittlement of cast duplex stainless steels under light-water operating conditions. Mechanical property data are presented from Charpy-impact, tensile, and J-R curve tests for several heats of cast stainless steel aged up to 10,000 h at 450, 400, 350, 320, and 290/sup 0/C. The results indicate that thermal aging increases the tensile strength and decreases the impact energy, J/sub IC/, and tearing modulus of the steels. Also, the ductile-to-brittle transition curve shifts to higher temperatures. The fracture toughness results are consistent with the Charpy-impact data, i.e., the relative reduction in J/sub IC/ is similar to the relative decrease in impact energy. The ferrite content and concentration of C in the steel have a strong effect on the overall process of low-temperature embrittlement. The low-carbon CF-3 steels are the most resistant and Mo-containing CF-8M steels are most susceptible to embrittlement. Weakening of the ferrite/austenite phase boundaries by carbide precipitates has a significant effect on the kinetics and extent of embrittlement of the high-carbon CF-8 and CF-8M steels, particularly after aging at temperatures greater than or equal to400/sup 0/C. The influence of N content and distribution of ferrite on loss of toughness are discussed. The data also indicate that existing correlations do not accurately represent the embrittlement behavior over the temperature range 280 to 450/sup 0/C, i.e., extrapolation of high-temperature data to reactor temperatures may not be valid for some compositions of cast stainless steel.
Date: June 1, 1987
Creator: Chopra, O.K. & Chung, H.M.
Partner: UNT Libraries Government Documents Department

High-temperature oxidation of Zircaloy in hydrogen-steam mixtures. [PWR; BWR]

Description: Oxidation rates of Zircaloy-4 cladding tubes have been measured in hydrogen-steam mixtures at 1200 to 1700/sup 0/C. For a given isothermal oxidation temperature, the oxide layer thicknesses have been measured as a function of time, steam supply rate, and hydrogen overpressure. The oxidation rates in the mixtures were compared with similar data obtained in pure steam and helium-steam environments under otherwise identical conditions. The rates in pure steam and helium-steam mixtures were equivalent and comparable to the parabolic rates obtained under steam-saturated conditions and reported in the literature. However, when the helium was replaced with hydrogen of equivalent partial pressure, a significantly smaller oxidation rate was observed. For high steam-supply rates, the oxidation kinetics in a hydrogen-steam mixture were parabolic, but the rate was smaller than for pure steam or helium-steam mixtures. Under otherwise identical conditions, the ratio of the parabolic rate for hydrogen-steam to that for pure steam decreased with increasing temperature and decreasing steam-supply rate.
Date: September 1, 1982
Creator: Chung, H.M. & Thomas, G.R.
Partner: UNT Libraries Government Documents Department

Microstructures of cast-duplex stainless steel after long-term aging

Description: Microstructures of cast-duplex stainless steels subjected to long-term aging either in the laboratory or during in-reactor service have been characterized and compared by TEM, SEM, and optical microscopy. The microstructural characteristics have been correlated with the impact failure behavior of the material. G-phase, ', and an unidentified Type X precipitate were responsible for the ferrite-phase embrittlement. Precipitation of M23C6 carbides on austenite-ferrite boundaries further degraded the reactor-aged material.
Date: October 1, 1985
Creator: Chung, H.M. & Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Long-term aging of cast stainless steels: Mechanisms and resulting properties

Description: Mechanical property data are presented from Charpy-impact, tensile, and J-R curve tests for several heats of cast stainless steel aged up to 10,000 h at 450, 400, 350, 320, and 290/sup 0/C. The results indicate that thermal aging increases the tensile strength and decreases the impactenergy, J/sub IC/ and tearing modulus of the steels. Also, the ductile-to-brittle transition curve shifts to higher temperatures. The low-carbon CF-3 steels were the most resistant and the molybdenum-containing high-carbon CF-8M steels were the most susceptible to low-temperature embrittlement. The influence of nitrogen content and distribution of ferrite on loss of toughness are discussed. Data also indicate that existing correlations do not accurately represent the embrittlement behavior over the temperature range 280 to 450/sup 0/C, i.e., extrapolation of high-temperature data to reactor temperatures may not be valid for some compositions of cast stainless steels. 13 refs., 13 figs., 2 tabs.
Date: September 1, 1987
Creator: Chopra, O.K. & Chung, H.M.
Partner: UNT Libraries Government Documents Department

Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

Description: This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain &gt;0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of &lt;0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of &lt;0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to the behavior of their high-C ...
Date: January 31, 2006
Creator: Chung, H. M.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Evaluation of aging of cast stainless steel components

Description: Cast stainless steel is used extensively in nuclear reactors for primary-pressure-boundary components such as primary coolant pipes, elbows, valves, pumps, and safe ends. These components are, however, susceptible to thermal aging embrittlement in light water reactors because of the segregation of Cr atoms from Fe and Ni by spinodal decomposition in ferrite and the precipitation of Cr-rich carbides on ferrite/austenite boundaries. A recent advance in understanding the aging kinetics is presented. Aging kinetics are strongly influenced by the synergistic effects of other metallurgical reactions that occur in parallel with spinodal decomposition, i.e., clustering of Ni, Mo, and Si solute atoms and the nucleation and growth of G-phase precipitates in the ferrite phase. A number of methods are outlined for estimating aging embrittlement under end-of-life of life-extension conditions, depending on several factors such as degree of permissible conservatism, availability of component archive material, and methods of estimating and verifying the activation energy of aging. 33 refs., 6 figs., 3 tabs.
Date: February 1, 1991
Creator: Chung, H.M.
Partner: UNT Libraries Government Documents Department

Fracture behavior of zircaloy spent-fuel cladding

Description: The Zircaloy cladding of water reactor fuel rods is susceptible to local breach-type failure, commonly known as pellet-cladding interaction (PCI) failure, during operational and off-normal power transients after the fuel has achieved a sufficiently high burnup. An optimization of power ramp procedures or fuel rod fabrication to minimize the cladding failure would result in a significant decrease in radiation exposure of plant personnel due to background and airborne radioactivity as well as an extension of core life in terms of allowable off-gas radioactivity. As part of a program to provide a better understanding of the fuel rod faiure phenomenon and to facilitate the formulation of a better failure criterion, a mechanistic study of the deformation and fracture behavior of high-burnup spent-fuel cladding is in progress under simulated PCI conditions.
Date: October 1, 1983
Creator: Chung, H.M.; Yaggee, F.L. & Kassner, T.F.
Partner: UNT Libraries Government Documents Department

Fracture behavior and microstructural characteristics of irradiated Zircaloy cladding

Description: Zircaloy cladding tube specimens from commercial power reactor fuel assemblies (burnup >22 MWd/kgU) have been deformed to fracture at 325/sup 0/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 11 irradiated specimens fractured by internal gas pressurization were examined by scanning electron microscopy, and 7 specimens were found to contain various degrees of the pseudocleavage feature that is characteristic of pellet-cladding interaction failures. Out of 10 test specimens fractured by expanding-mandrel loading, 5 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.
Date: June 1, 1985
Creator: Chung, H.M.; Yaggee, F.L. & Kassner, T.F.
Partner: UNT Libraries Government Documents Department

Fracture behavior of high-burnup spent-fuel cladding

Description: PCI-like, brittle-type failures, characterized by pseudocleavage-plus-fluting features in the fracture surface, branching cracks, and small diametral strain, were observed to occur at 292 to 325/sup 0/C in some batches of spent power-reactor fuel-cladding tubes under internal gas-pressurization and expanding-mandrel loading conditions in which the tests were not influenced by fission product simulants. Fractographic characteristics per se do not provide evidence for a PCI failure mechanism but should be deemed only as cooroborative in nature. Evaluation of TEM thin-foil specimens, obtained from regions adjacent to the brittle-type fracture sites, characteristically revealed extensive amounts of Zr/sub 3/O precipitates and a lack of slip dislocations. The precipitation of the Zr/sub 3/O phase appears to be enhanced by a high density of irradiation-induced defects. The brittle-type failure produced in the spent-fuel cladding tubes appears to be associated with segregation of oxygen to dislocation substructures and irradiation-induced defects, which leads to the formation of an ordered zirconium-oxygen phase of Zr/sub 3/O, an immobilization of dislocations, and minimal plastic deformation in the cladding material.
Date: October 1, 1983
Creator: Chung, H.M.; Yaggee, F.L. & Kassner, T.F.
Partner: UNT Libraries Government Documents Department

Correlation of microstructure and tensile and swelling behavior of neutron-irradiated vanadium alloys

Description: The microstructures of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were characterized by transmission electron microscopy (TEM) after neutron irradiation in the Fast Flux Test Facility (FFTF) at 420 and 600{degrees}C to influences up to 114 dpa. Two types of irradiation-induced precipitates were identified, i.e., Ti{sub 2}O and Ti{sub 5}(Si,P){sub 3}. Blocky Ti(O,N,C) precipitates, which form by thermal processes during ingot fabrication, also were observed in all unirradiated and irradiated specimens. Irradiation-induced precipitation of spherical (<15 nm in diameter) Ti{sub 5}(Si,P){sub 3} phase was associated with superior resistance to void swelling. In specimens with negligible swelling, Ti{sub 5}(Si,P){sub 3} precipitation was significant. It seems that ductility is significantly reduced when the precipitation of Ti{sub 2}O and Ti{sub 5}(Si,P){sub 3} is pronounced. These observations indicate that initial composition; fabrication processes; actual solute compositions of Ti, O, N, C, P, and Si after fabrication; O, N, and C uptake during service; and irradiation-induced precipitation ae interrelated and are important factors to consider in developing an optimized alloy. 15 refs., 8 figs.
Date: October 1, 1991
Creator: Chung, H.M. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

Description: Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.
Date: September 1, 1989
Creator: Chung, H.M. (Argonne National Lab., IL (USA))
Partner: UNT Libraries Government Documents Department