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The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

Description: Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pellet contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress effect. Kim et ...
Date: June 21, 2001
Creator: Kammenzind, B.F., Eklund, K.L. and Bajaj, R.
Partner: UNT Libraries Government Documents Department

Metallic and Non-Metallic Materials for the Primary Support Structure

Description: The primary support structure (PSS) is required for mechanical support of reactor module (RM) components and mounting of the RM to the spacecraft. The PSS would provide support and accept all loads associated with dynamic (e. g., launch and maneuvering) or thermally induced loading. Prior to termination of NRPCT involvement in Project Prometheus, the NRPCT Mechanical Systems team developed preliminary finite element models to gain a basic understanding of the behavior of the structure, but optimization of the models, specification of the final design, and materials selection were not completed. The Space Plant Materials team had evaluated several materials for potential use in the primary support structure, namely titanium alloys, beryllium, aluminum alloys and carbon-carbon composites. The feasibility of application of each material system was compared based on mass, stiffness, thermal expansion, and ease of fabrication. Due to insufficient data on environmental factors, such as temperatures and radiation, and limited modeling support, a final materials selection was not made.
Date: February 21, 2006
Creator: Wolf, RA & Corson, RP
Partner: UNT Libraries Government Documents Department

Modeling of Fission Gas Release in UO2

Description: A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].
Date: January 23, 2006
Creator: Krohn, MH
Partner: UNT Libraries Government Documents Department

MOLECULAR DYNAMICS SIMULATIONS OF DISPLACEMENT CASCADES IN MOLYBDENUM

Description: Molecular dynamics calculations have been employed to simulate displacement cascades in neutron irradiated Mo. A total of 90 simulations were conducted for PKA energies between 1 and 40 keV and temperatures from 298 to 923K. The results suggest very little effect of temperature on final defect count and configuration, but do display a temperature effect on peak defect generation prior to cascade collapse. Cascade efficiency, relative to the NRT model, is computed to lie between 1/4 and 1/3 in agreement with simulations performed on previous systems. There is a tendency for both interstitials and vacancies to cluster together following cascade collapse producing vacancy rich regions surrounded by interstitials. Although coming to rest in close proximity, the point defects comprising the clusters generally do not lie within the nearest neighbor positions of one another, except for the formation of dumbbell di-interstitials. Cascades produced at higher PKA energies (20 or 40 keV) exhibit the formation of subcascades.
Date: September 8, 2003
Creator: Smith, Richard Whiting
Partner: UNT Libraries Government Documents Department

On-Line Coolant Chemistry Analysis

Description: Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.
Date: July 19, 2006
Creator: Bachman, LM
Partner: UNT Libraries Government Documents Department

Processing of Refractory Metal Alloys for JOYO Irradiations

Description: This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang.
Date: February 21, 2006
Creator: Luther, RF & Petrichek, ME
Partner: UNT Libraries Government Documents Department

Quantitative Microanalysis with high Spatial Resolution: Application of FEG-DTEM XEDS Microanalysis to the Characterization of Complex Microstructures in Irradiated Low Alloy Steet

Description: To assist in the characterization of microstructural changes associated with irradiation damage in low alloy steels, the technique of quantitative x-ray mapping using a field emission gun scanning transmission electron microscope (FEG-STEM) equipped with an x-ray energy Dispersive spectrometer (XEDS) has been employed. Quantitative XEDS microanalyses of the matrix and grain boundaries of irradiated specimens have been compared with previous quantitative analyses obtained using 3D-Atom Probe Field-Ion Microscopy (3D-APFIM). In addition, the FEG-STEM XEDS maps obtained from the irradiated steel have revealed the presence of 2 to 3 nm Ni-enriched 'precipitates' in the matrix, which had previously been detected using 3D-APFIM. These quantitative FEG-STEM XEDS results represent the first direct and independent microchemical corroboration of the 3D-APFIM results showing ultra-fine irradiation-induced hardening features in low alloy steel.
Date: November 14, 2001
Creator: Williams, D.B., Watanabe, M. and Burke, M.G.
Partner: UNT Libraries Government Documents Department

A Review of Tribological Coatings for Control Drive Mechanisms in Space Reactors

Description: Tribological coatings must provide lubrication for moving components of the control drive mechanism for a space reactor and prevent seizing due to friction or diffusion welding to provide highly reliable and precise control of reflector position over the mission lifetime. Several coatings were evaluated based on tribological performance at elevated temperatures and in ultrahigh vacuum environments. Candidates with proven performance in the anticipated environment are limited primarily to disulfide materials. Irradiation data for these coatings is nonexistent. Compatibility issues between coating materials and structural components may require the use of barrier layers between the solid lubricant and structural components to prevent deleterious interactions. It would be advisable to consider possible lubricant interactions prior to down-selection of structural materials. A battery of tests was proposed to provide the necessary data for eventual solid lubricant/coating selection.
Date: February 21, 2006
Creator: Larkin, CJ; Edington, JD & Close, BJ
Partner: UNT Libraries Government Documents Department

Stress Corrosion Cracking Response of 304 Stainless Steel in ASerated and Dearated Water

Description: Scoping stress corrosion cracking (SCC) tests of 304 stainless steel (SS) were performed in 75 C and 250 C aerated pressurized water (APW) and 250 C deaerated pressurized water (DPW). The 250 C APW environment was used to initiate intergranular stress corrosion cracking (IGSCC) and then the water was deaerated and hydrogenated to see if IGSCC continued in 250 C DPW. Tests were performed with and without 200 ppb SO{sub 4}{sup =}. The 304 SS test materials were evaluated in either the as-received, heavily sensitized (649 C for 1 h), fully sensitized (1099 C for 1 h/water quench/621 C for 17 h) or 20% cold rolled condition. At the beginning of each test sequence, specimens were subjected to continuous cycling with a 500s rise/500s fall or a 5000s rise/500s fall to promote the transition from a transgranular (TG) precrack to an IG crack. After generating a uniform crack under continuous cycling conditions, a trapezoidal waveform with 500s rise/9000s hold/500s fall was used to characterize the SCC behavior. Crack growth rates (CGRs) were monitored continuously with the electric potential drop (EPD) method and were corrected based on physical crack length measurements obtained when specimens were destructively evaluated. Continuous cycling with a 500s or 5000s rise time was found to produce both TG faceting and IGSCC in fully sensitized 304 SS tested in 75 C APW with 7 ppm O{sub 2} and 200 ppb SO{sub 4}{sup =}. However, no measurable crack extension occurred when a 9000 s hold time was introduced. Extensive IGSCC occurred in heavily sensitized and fully sensitized 304 SS in 250 C APW with 1 ppm O{sub 2} and 200 ppb SO{sub 4}{sup =}. IGSCC initiated under continuous cycling conditions with a 500 s rise time, and rapid IGSCC occurred when a 9000 s hold time was introduced. During ...
Date: April 30, 2007
Creator: Mills, W. J.
Partner: UNT Libraries Government Documents Department

Summary of Dissimilar Metal Joining Trials Conducted by Edison Welding Institute

Description: Under the direction of the NASA-Glenn Research Center, the Edison Welding Institute (EWI) in Columbus, OH performed a series of non-fusion joining experiments to determine the feasibility of joining refractory metals or refractory metal alloys to Ni-based superalloys. Results, as reported by EWI, can be found in the project report for EWI Project 48819GTH (Attachment A, at the end of this document), dated October 10, 2005. The three joining methods used in this investigation were inertia welding, magnetic pulse welding, and electro-spark deposition joining. Five materials were used in these experiments: Mo-47Re, T-111, Hastelloy X, Mar M-247 (coarse-grained, 0.5 mm to several millimeter average grain size), and Mar M-247 (fine-grained, approximately 50 {micro}m average grain size). Several iterative trials of each material combination with each joining method were performed to determine the best practice joining method. Mo-47Re was found to be joined easily to Hastelloy X via inertia welding, but inertia welding of the Mo-alloy to both Mar M-247 alloys resulted in inconsistent joint strength and large reaction layers between the two metals. T-111 was found to join well to Hastelloy X and coarse-grained Mar M-247 via inertia welding, but joining to fine-grained Mar M-247 resulted in low joint strength. Magnetic pulse welding (MPW) was only successful in joining T-111 tubing to Hastelloy X bar stock. The joint integrity and reaction layer between the metals were found to be acceptable. This single joining trial, however, caused damage to the electromagnetic concentrators used in this process. Subsequent design efforts to eliminate the problem resulted in a loss of power imparted to the accelerating work piece, and results could not be reproduced. Welding trials of Mar M-247 to T-111 resulted in catastrophic failure of the bar stock, even at lower power. Electro-spark deposition joining of Mo-47Re, in which the deposited material was ...
Date: November 18, 2005
Creator: Lambert, MJ
Partner: UNT Libraries Government Documents Department

Testing Results of Magnetostrictive Ultrasonic Sensor Cables for Signal Loss

Description: The purpose of this test was to determine the signal strength and resolution losses of a magnetostrictive ultrasonic system with an extended signal cable. The cable of interest carries electrical signals between the pulse generator/receiver and the magnetostrictive transducer. It was desired to determine the loss introduced by different lengths of the signal cable (6', 100', and 200').
Date: May 1, 2005
Creator: Evans, JT
Partner: UNT Libraries Government Documents Department

Transmission electron microscopy of oxide dispersion strengthened (ODS) molybdenum: effects of irradiation on material microstructure

Description: Oxide dispersion strengthened (ODS) molybdenum has been characterized using transmission electron microscopy (TEM) to determine the effects of irradiation on material microstructure. This work describes the results-to-date from TEM characterization of unirradiated and irradiated ODS molybdenum. The general microstructure of the unirradiated material consists of fine molybdenum grains (< 5 {micro}m average grain size) with numerous low angle boundaries and isolated dislocation networks. 'Ribbon'-like lanthanum oxides are aligned along the working direction of the product form and are frequently associated with grain boundaries, serving to inhibit grain boundary and dislocation movement. In addition to the 'ribbons', discrete lanthanum oxide particles have also been detected. After irradiation, the material is characterized by the presence of nonuniformly distributed large ({approx} 20 to 100 nm in diameter), multi-faceted voids, while the molybdenum grain size and oxide morphology appear to be unaffected by irradiation.
Date: March 3, 2003
Creator: Baranwal, R. and Burke, M.G.
Partner: UNT Libraries Government Documents Department

Double Retort System for Materials Compatibility Testing

Description: With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the Space Nuclear Power Plant (SNPP) for Project Prometheus (References a and b) there was a need to investigate compatibility between the various materials to be used throughout the SNPP. Of particular interest was the transport of interstitial impurities from the nickel-base superalloys, which were leading candidates for most of the piping and turbine components to the refractory metal alloys planned for use in the reactor core. This kind of contamination has the potential to affect the lifetime of the core materials. This letter provides technical information regarding the assembly and operation of a double retort materials compatibility testing system and initial experimental results. The use of a double retort system to test materials compatibility through the transfer of impurities from a source to a sink material is described here. The system has independent temperature control for both materials and is far less complex than closed loops. The system is described in detail and the results of three experiments are presented.
Date: February 23, 2006
Creator: Munne, V. & Carelli, E. V.
Partner: UNT Libraries Government Documents Department

THE EFFECT OF INTERSTITIAL N ON GRAIN BOUNDARY COHESIVE STRENGTH IN Fe

Description: Increased nitrogen levels have been correlated with decreased ductility and elevated ductile-to-brittle transition temperature in pressure vessel steels [1]. However, the exact role played by nitrogen in the embrittlement of steels remains unclear. Miller and Burke have reported atom probe ion microscopy findings from neutron-irradiated low-alloy pressure vessel steel showing the presence of a 1 to 2 ruonolayer thick film of Mo, N, and C at prior austenitic grain boundaries (GB's) [2], suggesting a role for nitrogen as an intergranular embrittler. It is of interest for the development of mitigation strategies whether nitrogen must combine with other impurities to form nitride precipitates in order to exert an embrittling effect. Briant et al [1] have associated the embrittling effect of N in steels exclusively with intergranular nitride formation. This association suggests that high nitrogen levels may be acceptable if nitride precipitation at grain boundaries is suppressed. To address whether precipitate formation is indeed essential to the N embrittlement process in pressure vessel steel, a computational study was undertaken to ascertain whether the presence of interstitial nitrogen alone could embrittle an Fe GB. If so, nitrogen in any form must be kept completely away from the grain boundaries, if not out of the material altogether. The effect of interstitial N on the cohesion of an Fe {Sigma}3[110](111) grain boundary (GB) was investigated by ab-initio electronic structure calculations to reveal that free interstitial N produces a large strengthening energy, reduces the magnetic moments of the GB Fe atoms and is embrittling at the GB's.
Date: September 22, 2003
Creator: Miyoung, Kim; Gellar, Clint B. & Freeman, A. F.
Partner: UNT Libraries Government Documents Department

EFFECT OF UNBROKEN LIGAMENTS ON STRESS CORROSION CRACKING BEHAVIOR OF ALLOY 82H WELDS

Description: Previously reported stress corrosion cracking (SCC) rates for Alloy 82H gas-tungsten-arc welds tested in 360 C water showed tremendous variability. The excessive data scatter was attributed to the variations in microstructure, mechanical properties and residual stresses that are common in welds. In the current study, however, re-evaluation of the SCC data revealed that the large data scatter was an anomaly due to erroneous crack growth rates inferred from crack mouth opening displacement (CMOD) measurements. Apparently, CMOD measurements provided reasonably accurate SCC rates for some specimens, but grossly overestimated rates in others. The overprediction was associated with large unbroken ligaments that often form in welds in the wake of advancing crack fronts. When ligaments were particularly large, they prevented crack mouth deflection, so apparent crack incubation times (i.e. period of time before crack advance commences) based on CMOD measurements were unrealistically long. During the final states of testing, ligaments began to separate allowing the crack mouth to open rather quickly. This behavior was interpreted as a rapid crack advance, but it actually reflects the ligament separation rate, not the SCC rate. Revised crack growth rates obtained in this study exhibit substantially less scatter than that previously reported. The effects of crack orientation and fatigue flutter loading on SCC rates in 82H welds are also discussed.
Date: February 20, 2003
Creator: Mills, W.J. and Brown, C.M.
Partner: UNT Libraries Government Documents Department

Evaluation of Irradiation Embrittlement of A508 Gr 4N and Comparison to Other Low-Alloy Steels

Description: A508 Gr 4N has improved fracture toughness because of the addition of 3% nickel, compared to typical low alloy steels which have less than 1% nickel. However, there is an expectation in much of the recent literature, based mostly on low-alloy steels with nickel below 1%, that irradiation embrittlement will increase with increasing nickel (Ni) content. In contrast, the raw irradiation test data show that ASTM A508 Grade 4N containing up to 3.7% nickel, 0.1% Cu and 0.01% P does not show enhanced irradiation embrittlement. A simple statistical fit to irradiation dose and irradiation temperature was developed to make direct comparisons to other low-alloy steels. Since the A508 Gr 4N data showed little discernible effect of Cu in the raw data, the damage may be classified as 'matrix' damage. The peak irradiation embrittlement of A508 Gr 4N is no greater than that of A508 Gr 2, a 0.7% Ni forging material tested under similar conditions with similar limits on Cu and P. At high dose (80 mdpa) the average embrittlement of A508 Gr 4N is slightly higher (33%) than the lower nickel materials. This trend also occurs for low copper A533B and A302B plate material. The irradiation temperature dependence of embrittlement in A508 Gr 4N is nearly the same as other low copper low-alloy steels tested over a wide range of temperatures. The increase in Charpy transition temperature in A508 Gr 4N is due to radiation hardening, and the ratio of TTS to yield strength increase in 3 Ni steels is nearly identical to that observed for conventional low-alloy steels with lower nickel. A very detailed statistical fit was made to the overall data on A508 Gr 4N to evaluate the sensitivity of embrittlement to minor elements and to compare to results from the US surveillance test data, which is ...
Date: June 17, 2002
Creator: G.L. Wire, W. J. Beggs and T.R. Leax
Partner: UNT Libraries Government Documents Department

Experimental Design for Evaluation of Co-extruded Refractory Metal/Nickel Base Superalloy Joints

Description: Prior to the restructuring of the Prometheus Program, the NRPCT was tasked with delivering a nuclear space reactor. Potential NRPCT nuclear space reactor designs for the Prometheus Project required dissimilar materials to be in contact with each other while operating at extreme temperatures under irradiation. As a result of the high reactor core temperatures, refractory metals were the primary candidates for many of the reactor structural and cladding components. They included the tantalum-base alloys ASTAR-811C and Ta-10W, the niobium-base alloy FS-85, and the molybdenum base alloys Moly 41-47.5 Rhenium. The refractory metals were to be joined to candidate nickel base alloys such as Haynes 230, Alloy 617, or Nimonic PE 16 either within the core if the nickel-base alloys were ultimately selected to form the outer core barrel, or at a location exterior to the core if the nickel-base alloys were limited to components exterior to the core. To support the need for dissimilar metal joints in the Prometheus Project, a co-extrusion experiment was proposed. There are several potential methods for the formation of dissimilar metal joints, including explosive bonding, friction stir welding, plasma spray, inertia welding, HIP, and co-extrusion. Most of these joining methods are not viable options because they result in the immediate formation of brittle intermetallics. Upon cooling, intermetallics form in the weld fusion zone between the joined metals. Because brittle intermetallics do not form during the initial bonding process associated with HIP, co-extrusion, and explosive bonding, these three joining procedures are preferred for forming dissimilar metal joints. In reference to a Westinghouse Astronuclear Laboratory report done under a NASA sponsored program, joints that were fabricated between similar materials via explosive bonding had strengths that were directly affected by the width of the diffusion barrier. It was determined that the diffusion zone should not exceed a critical ...
Date: December 16, 2005
Creator: Petrichek, ME
Partner: UNT Libraries Government Documents Department

Experimental Evaluation of Tude Support Plate Crevice Chemistry

Description: A test methodology for measuring temperature, impedance, pH, and electrochemical potential distributions within a sludge-packed tube support plate crevice in a laboratory test is described. The method successfully showed that there were large concentration gradients between the tube and tube support plate sides of the crevice. The testing also showed that strong bases concentrated more effectively than strong acids, and that the crevice pH, when exposed to seawater-based solutions, increased with increasing superheat and decreasing bulk concentration. The large variations in the crevice chemistry observed under heat transfer were eliminated upon shutdown. These new test data suggest that it might be beneficial to evaluate the variation in the extent of stress corrosion cracking with tube support plate elevation found in some steam generators in light of local chemistry changes, as well as the variation in tubing temperature. Because of the large crevice chemistry gradients during boiling heat transfer and their subsequent homogenization upon test shutdown, the results suggest reassessing the use of hideout return measurements and tube deposit analyses in industry to infer the crevice chemistry under heat transfer conditions.
Date: May 8, 2001
Creator: Baum, Allen
Partner: UNT Libraries Government Documents Department

Failure Analysis of Cracked FS-85 Tubing and ASTAR-811C End Caps

Description: Failure analyses were performed on cracked FS-85 tubing and ASTAR-811C and caps which had been fabricated as components of biaxial creep specimens meant to support materials testing for the NR Space program. During the failure analyses of cracked FS-85 tubing, it was determined that the failure potentially could be due to two effects: possible copper contamination from the EDM (electro-discharge machined) recast layer and/or an insufficient solution anneal. to prevent similar failures in the future, a more formal analysis should be done after each processing step to ensure the quality of the material before further processing. During machining of the ASTAR-811FC rod to form end caps for biaxial creep specimens, linear defects were observed along the center portion of the end caps. These defects were only found in material that was processed from the top portion of the ingot. The linear defects were attributed to a probable residual ingot pipe that was not removed from the ingot. During the subsequent processing of the ingot to rod, the processing temperatures were not high enough to allow self healing of the ingot's residual pipe defect. To prevent this from occurring in the future, it is necessary to ensure that complete removal of the as-melted ingot pipe is verified by suitable non-destructive evaluation (NDE).
Date: February 9, 2006
Creator: Petrichek, ME
Partner: UNT Libraries Government Documents Department

FRACTURE BEHAVIOR OF ALLOY 600, ALLOY 690, EN82H WELDS AND EN52 WELDS IN WATER

Description: The cracking resistance of Alloy 600, Alloy 690 and their welds, EN82H and EN52, was characterized by conducting J{sub IC} rising load tests in air and hydrogenated water and cooldown testing in water under constant-displacement conditions. All test materials displayed excellent toughness in air and high temperature water, but Alloy 690 and the two welds were severely embrittled in low temperature water. In 54 C water with 150 cc H{sub 2}/kg H{sub 2}O, J{sub IC} values were reduced by 70% to 95%, relative to their air counterpart. The toughness degradation was associated with a fracture mechanism transition from microvoid coalescence to intergranular fracture. Comparison of the cracking response in water with that for hydrogen-precharged specimens tested in air demonstrated that susceptibility to low temperature crack propagation (LTCP) is due to hydrogen embrittlement of grain boundaries. The effects of water temperature, hydrogen content and loading rate on LTCP were studied. In addition, testing of specimens containing natural weld defects and as-machined notches was performed to determine if low temperature cracking can initiate at these features. Unlike the other materials, Alloy 600 is not susceptible to LTCP as the toughness in 54 C water remained high and a microvoid coalescence mechanism was operative in both air and water. Cooldown testing of EN82H welds under constant-displacement conditions was performed to determine if LTCP data from rising load J{sub IC}/K{sub Pmax} tests predict the onset of LTCP for other load paths. In these tests, bolt-loaded CT specimens were subjected to 288 C water for up to 1 week, cooled to 54 C and held in 54 C hydrogenated water for 1 week. This cycle was repeated up to 6 times. For two of the three welds tested, critical K{sub I} levels for LTCP under constant-displacement conditions were much higher than rising load K{sub Pmax} ...
Date: January 11, 2000
Creator: Mills, W.J., Brown, C.M. and Burke, M.G.
Partner: UNT Libraries Government Documents Department

Fuel System Compatibility Issues for Prometheus-1

Description: Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.
Date: January 20, 2006
Creator: Noe, DC; Gibbard, KB & Krohn, MH
Partner: UNT Libraries Government Documents Department