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A Hierarchical Upscaling Method for Predicting Strength of Materials under Thermal, Radiation and Mechanical loading - Irradiation Strengthening Mechanisms in Stainless Steels

Description: Stainless steels based on Fe-Cr-Ni alloys are the most popular structural materials used in reactors. High energy particle irradiation of in this kind of polycrystalline structural materials usually produces irradiation hardening and embrittlement. The development of predictive capability for the influence of irradiation on mechanical behavior is very important in materials design for next-generation reactors. Irradiation hardening is related to structural information crossing different length scale, such as composition, dislocation, crystal orientation distribution and so on. To predict the effective hardening, the influence factors along different length scales should be considered. A multiscale approach was implemented in this work to predict irradiation hardening of iron based structural materials. Three length scales are involved in this multiscale model: nanometer, micrometer and millimeter. In the microscale, molecular dynamics (MD) was utilized to predict on the edge dislocation mobility in body centered cubic (bcc) Fe and its Ni and Cr alloys. On the mesoscale, dislocation dynamics (DD) models were used to predict the critical resolved shear stress from the evolution of local dislocation and defects. In the macroscale, a viscoplastic self-consistent (VPSC) model was applied to predict the irradiation hardening in samples with changes in texture. The effects of defect density and texture were investigated. Simulated evolution of yield strength with irradiation agrees well with the experimental data of irradiation strengthening of stainless steel 304L, 316L and T91. This multiscale model we developed in this project can provide a guidance tool in performance evaluation of structural materials for next-generation nuclear reactors. Combining with other tools developed in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the models developed will have more impact in improving the reliability of current reactors and affordability of new reactors.
Date: July 1, 2011
Creator: Li, Dongsheng; Zbib, Hussein M.; Garmestani, Hamid; Sun, Xin & Khaleel, Mohammad A.
Partner: UNT Libraries Government Documents Department


Description: A high temperature guarded-comparative-longitudinal heat flow measurement system has been built to measure the thermal conductivity of a composite nuclear fuel compact. It is a steady-state measurement device designed to operate over a temperature range of 300 K to 1200 K. No existing apparatus is currently available for obtaining the thermal conductivity of the composite fuel in a non-destructive manner due to the compact’s unique geometry and composite nature. The current system design has been adapted from ASTM E 1225. As a way to simplify the design and operation of the system, it uses a unique radiative heat sink to conduct heat away from the sample column. A finite element analysis was performed on the measurement system to analyze the associated error for various operating conditions. Optimal operational conditions have been discovered through this analysis and results are presented. Several materials have been measured by the system and results are presented for stainless steel 304, inconel 625, and 99.95% pure iron covering a range of thermal conductivities of 10 W/m*K to 70 W/m*K. A comparison of the results has been made to data from existing literature.
Date: March 1, 2011
Creator: Phillips, Jeff; Jensen, Colby; Xing, Changhu & Ban, Heng
Partner: UNT Libraries Government Documents Department

Thermal Properties of Structural Materials Found in Light Water Reactor Vessels

Description: High temperature material property data for structural materials used in existing Light Water Reactors (LWRs) are limited. Often, extrapolated values recommended in the literature differ significantly. To reduce such uncertainties, new data for SA533 Grade B, Class 1 (SA533B1) low alloy steel, Stainless Steel 304 (SS304), and Inconel 600, found in Light Water Reactor (LWR) vessels and penetrations, were acquired and tested using material property systems available at the High Temperature Test Laboratory (HTTL) at the Idaho National Laboratory (INL). Properties measured include thermal expansion, specific heat capacity, and thermal diffusivity for temperatures up to 1200 oC. From these results, thermal conductivity and density were calculated. Results show that, in some cases, previously recommended values for these material differ significantly from measured values at high temperatures. This is especially true for SA533B1, as previous data do not account for the phase transformation of this material between 740 oC and 840 oC.
Date: November 1, 2009
Creator: Daw, J. E.; Rempe, J. L. & Knudson, D. L.
Partner: UNT Libraries Government Documents Department

Criticality Evaluation of Plutonium-239 Moderated by High-Density Polyethylene in Stainless Steel and Aluminum Containers Suitable for Non-Exclusive Use Transport

Description: Research is conducted at the Joint Actinide Shock Physics Experimental Facility (JASPER) on the effects of high pressure and temperature environments on plutonium-239, in support of the stockpile stewardship program. Once an experiment has been completed, it is necessary to transport the end products for interim storage or final disposition. Federal shipping regulations for nonexclusive use transportation require that no more than 180 grams of fissile material are present in at least 360 kilograms of contiguous non-fissile material. To evaluate the conservatism of these regulatory requirements, a worst-case scenario of 180g {sup 239}Pu and a more realistic scenario of 100g {sup 239}Pu were modeled using one of Lawrence Livermore National Laboratory's Monte Carlo transport codes known as COG 10. The geometry consisted of {sup 239}Pu spheres homogeneously mixed with high-density polyethylene surrounded by a cube of either stainless steel 304 or aluminum. An optimized geometry for both cube materials and hydrogen-to-fissile isotope (H/X) ratio were determined for a single unit. Infinite and finite 3D arrays of these optimized units were then simulated to determine if the systems would exceed criticality. Completion of these simulations showed that the optimal H/X ratio for the most reactive units ranged from 800 to 1600. A single unit of either cube type for either scenario would not reach criticality. An infinite array was determined to reach criticality only for the 180g case. The offsetting of spheres in their respective cubes was also considered and showed a considerable decrease in the number of close-packed units needed to reach criticality. These results call into question the current regulations for fissile material transport, which under certain circumstances may not be sufficient in preventing the development of a critical system. However, a conservative, theoretical approach was taken in all assumptions and such idealized configurations may not be likely to ...
Date: August 10, 2007
Creator: Watson, T T
Partner: UNT Libraries Government Documents Department

On the interface between LENS deposited stainless steel 304L repair geometry and cast or machined components.

Description: Laser Engineered Net Shaping (LENS) is being evaluated for use as a metal component repair/modification process for the NWC. An aspect of the evaluation is to better understand the characteristics of the interface between LENS deposited material and the substrate on which it is deposited. A processing and metallurgical evaluation was made on LENS processed material fabricated for component qualification tests. A process parameter evaluation was used to determine optimum build parameters and these parameters were used in the fabrication of tensile test specimens to study the characteristics of the interface between LENS deposited material and several types of substrates. Analyses of the interface included mechanical properties, microstructure, and metallurgical integrity. Test samples were determined for a variety of geometric configurations associated with interfaces between LENS deposited material and both wrought base material and previously deposited LENS material. Thirteen different interface configurations were fabricated for evaluation representing a spectrum of deposition conditions from complete part build, to hybrid substrate-LENS builds, to repair builds for damaged or re-designed housings. Good mechanical properties and full density were observed for all configurations. When tested to failure, fracture occurred by ductile microvoid coalescence. The repair and hybrid interfaces showed the same metallurgical integrity as, and had properties similar to, monolithic LENS deposits.
Date: December 1, 2004
Creator: Smugeresky, John E.; Harris, Marc F.; Griffith, Michelle Lynn; Gill, David Dennis & Robino, Charles Victor
Partner: UNT Libraries Government Documents Department

Temperature effects on the mechanical properties of annealed and HERF 304L stainless steel.

Description: The effect of temperature on the tensile properties of annealed 304L stainless steel and HERF 304L stainless steel forgings was determined by completing experiments over the moderate range of -40 F to 160 F. Temperature effects were more significant in the annealed material than the HERF material. The tensile yield strength of the annealed material at -40 F averaged twenty two percent above the room temperature value and at 160 F averaged thirteen percent below. The tensile yield strength for the three different geometry HERF forgings at -40 F and 160 F changed less than ten percent from room temperature. The ultimate tensile strength was more temperature dependent than the yield strength. The annealed material averaged thirty six percent above and fourteen percent below the room temperature ultimate strength at -40 F and 160 F, respectively. The HERF forgings exhibited similar, slightly lower changes in ultimate strength with temperature. For completeness and illustrative purposes, the stress-strain curves are included for each of the tensile experiments conducted. The results of this study prompted a continuation study to determine tensile property changes of welded 304L stainless steel material with temperature, documented separately.
Date: November 1, 2004
Creator: Antoun, Bonnie R.
Partner: UNT Libraries Government Documents Department

Waste Package and Material Testing for the Proposed Yucca Mountain High Level Waste Repository

Description: Over the repository lifetime, the waste package containment barriers will perform various functions that will change with time. During the operational period, the barriers will function as vessels for handling, emplacement, and waste retrieval (if necessary). During the years following repository closure, the containment barriers will be relied upon to provide substantially complete containment, through 10,000 years and beyond. Following the substantially complete containment phase, the barriers and the waste package internal structures help minimize release of radionuclides by aqueous- and gaseous-phase transport. These requirements have lead to a defense-in-depth design philosophy. A multi-barrier design will result in a lower breach rate distributed over a longer period of time, thereby ensuring the regulatory requirements are met. The design of the Engineered Barrier System (EBS) has evolved. The initial waste package design was a thin walled package, 3/8 inch of stainless steel 304, that had very limited capacity, (3 PWR and 4 BWR assemblies) and performance characteristics, 300 to 1,000 years. This design required over 35,000 waste packages compared to today's design of just over 10,000 waste packages. The waste package designs are now based on a defense-in-depth/multi-barrier philosophy and have a capacity similar to the standard storage and rail transported spent nuclear fuel casks. Concurrent with the development of the design of the waste packages, a comprehensive waste package materials testing program has been undertaken to support the selection of containment barrier materials and to develop predictive models for the long-term behavior of these materials under expected repository conditions. The testing program includes both long-term and short-term tests and the results from these tests combination with the data published in the open literature are being used to develop models for predicting performance of the waste packages.
Date: January 1, 2002
Creator: Doering, Thomas & Pasupathi, V.
Partner: UNT Libraries Government Documents Department

Tensile stress corrosion cracking of type 304 stainless steel irradiated to very high dose

Description: Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20--100 displacement per atom or dpa) by the end of life. The data bases and mechanistic understanding of, the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-commotion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-11 reactor after irradiation to {approximately}50 dpa at {approximately}370 C. Slow-strain-rate tensile tests were conducted at 289 C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microcopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.
Date: September 1, 2001
Creator: Chung, H. M.; Ruther, W. E.; Strain, R. V. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

Material Corrosion and Plate-Out Test of Types 304L and 316L Stainless Steel

Description: Corrosion and plate-out tests were performed on 304L and 316L stainless steel in pretreated Envelope B and Envelope C solutions. Flat coupons of the two stainless steels were exposed to 100 degrees C liquid and to 74 degrees C and 88 degrees C vapor above the solutions for 61 days. No significant corrosion was observed either by weight-loss measurements or by microscopic examination. Most coupons had small weight gains due to plate-out of solids, which remained to some extent even after 24-hour immersion in 1 N nitric acid at room temperature. Plate-out was more significant in the Envelope B coupons, with film thickness from less than 0.001 in. to 0.003-inches.
Date: February 6, 2001
Creator: Zapp, P.E.
Partner: UNT Libraries Government Documents Department

Active Waste Materials Corrosion and Decontamination Tests

Description: Stainless steel alloys, 304L and 316L, were corrosion tested in representative radioactive samples of three actual Hanford tank waste solutions (Tanks AW-101, C-104, AN-107). Both the 304L and 316L exhibited good corrosion performance when immersed in boiling waste solutions. The maximum general corrosion rate was 0.015 mm/y (0.60 mils per year). Generally, the 304L had a slightly higher rate than the 316L. No localized attack was observed after 122 days of testing in the liquid phase, liquid/vapor phase, or vapor phase. Radioactive plate-out decontamination tests indicated that a 24-hour exposure to 1 {und M} HNO{sub 3} could remove about 99% of the radioactive components in the metal film when exposed to the C-104 and AN-107 solutions. The decontamination results are less certain for the AW-101 solution, since the initial contamination readings exceeded the capacity of the meter used for this test.
Date: August 15, 2000
Creator: Danielson, MJ; Elmore, MR & Pitman, SG
Partner: UNT Libraries Government Documents Department

Microstructures of laser deposited 304L austenitic stainless steel

Description: Laser deposits fabricated from two different compositions of 304L stainless steel powder were characterized to determine the nature of the solidification and solid state transformations. One of the goals of this work was to determine to what extent novel microstructure consisting of single-phase austenite could be achieved with the thermal conditions of the LENS [Laser Engineered Net Shape] process. Although ferrite-free deposits were not obtained, structures with very low ferrite content were achieved. It appeared that, with slight changes in alloy composition, this goal could be met via two different solidification and transformation mechanisms.
Date: May 22, 2000
Partner: UNT Libraries Government Documents Department

Residual stress determination from a laser-based curvature measurement

Description: Thermally sprayed coating characteristics and mechanical properties are in part a result of the residual stress developed during the fabrication process. The total stress state in a coating/substrate is comprised of the quench stress and the coefficient of thermal expansion (CTE) mismatch stress. The quench stress is developed when molten particles impact the substrate and rapidly cool and solidify. The CTE mismatch stress results from a large difference in the thermal expansion coefficients of the coating and substrate material. It comes into effect when the substrate/coating combination cools from the equilibrated deposit temperature to room temperature. This paper describes a laser-based technique for measuring the curvature of a coated substrate and the analysis required to determine residual stress from curvature measurements. Quench stresses were determined by heating the specimen back to the deposit temperature thus removing the CTE mismatch stress. By subtracting the quench stress from the total residual stress at room temperature, the CTE mismatch stress was estimated. Residual stress measurements for thick (>1mm) spinel coatings with a Ni-Al bond coat on 304 stainless steel substrates were made. It was determined that a significant portion of the residual stress results from the quenching stress of the bond coat and that the spinel coating produces a larger CTE mismatch stress than quench stress.
Date: May 8, 2000
Creator: Swank, W. D.; Gavalya, R. A.; Wright, J. K. & Wright, R. N.
Partner: UNT Libraries Government Documents Department

Examination of irradiated 304L stainless steel to 6061-T6 aluminum inertia welded transition joints after irradiation in a spallation neutron

Description: The Savannah River Technology Center (SRTC) designed and fabricated tritium target/blanket assemblies which were irradiated for six months at the Los Alamos Neutron Science Center (LANSCE). Cooling water was supplied to the assemblies through 1 inch diameter 304L Stainless Steel (SS) tubing. To attach the 304L SS tubing to the modules a 304L SS to 6061-T6 Aluminum (Al) inertia welded transition joint was used. These SS/Al inertia weld transition joints simulate expected transition joints in the Accelerator Production of Tritium (APT) Target/Blanket where as many as a thousand SS/Al weld transition joints will be used. Materials compatibility between the 304L SS and the 6061-T6 Al in the spallation neutron environment is a major concern as well as the corrosion associated with the cooling water flowing through the piping. The irradiated inertia weld examination will be discussed.
Date: April 28, 2000
Creator: Dunn, K.A.
Partner: UNT Libraries Government Documents Department

Residual stress measurement with high energy x-rays at the Advanced Photon Source.

Description: Preliminary measurements with high energy x-rays from the SRI CAT 1-ID beam line at the Advanced Photon show great promise for the measurement of stress and strain using diffraction. Comparisons are made with neutron measurements. Measurements of strains in a 2 mm thick 304 stainless steel weld show that excellent strain and spatial resolutions are possible. With 200 {micro}m slits, strain resolutions of 1 x 10{sup {minus}5} were achieved.
Date: March 2, 2000
Creator: Winholtz, R. A.; Haeffner, D. R.; Green, R.E.L.; Varma, R. & Hammond, D.
Partner: UNT Libraries Government Documents Department

Corrosion of candidate container materials by Yucca Mountain bacteria

Description: Several candidate container materials have been studied in modified Yucca Mountain (YM) ground water in the presence or absence of YM bacteria. YM bacteria increased corrosion rates by 5-6 fold in UNS G10200 carbon steel, and nearly 100-fold in UNS NO4400 Ni-Cu alloy. YM bacteria caused microbiologically influenced corrosion (MIC) through de-alloying or Ni-depletion of Ni-Cu alloy as evidenced by scanning electronic microscopy (SEM) and inductively coupled plasma spectroscopy (ICP) analysis. MIC rates of more corrosion-resistant alloys such as UNS NO6022 Ni-Cr- MO-W alloy, UN's NO6625 Ni-Cr-Mo alloy, and UNS S30400 stainless steel were measured below 0.05 umyr, however YM bacteria affected depletion of Cr and Fe relative to Ni in these materials. The chemical change on the metal surface caused by depletion was characterized in anodic polarization behavior. The anodic polarization behavior of depleted Ni-based alloys was similar to that of pure Ni. Key words: MIC, container materials, YM bacteria, de-alloying, Ni-depletion, Cr-depletion, polarization resistance, anodic polarization,
Date: December 10, 1999
Creator: Horn, J; Jones, D; Lian, T; Martin, S & Rivera, A
Partner: UNT Libraries Government Documents Department

Analysis and validation of laser spot weld-induced distortion

Description: Laser spot welding is an ideal process for joining small parts with tight tolerances on weld size, location, and distortion, particularly those with near-by heat sensitive features. It is also key to understanding the overlapping laser spot seam welding process. Rather than attempting to simulate the laser beam-to-part coupling (particularly if a keyhole occurs), it was measured by calorimetry. This data was then used to calculate the thermal and structural response of a laser spot welded SS304 disk using the finite element method. Five combinations of process parameter values were studied. Calculations were compared to experimental data for temperature and distortion profiles measured by thermocouples and surface profiling. Results are discussed in terms of experimental and modeling factors. The authors then suggest appropriate parameters for laser spot welding.
Date: December 9, 1999
Creator: Knorovsky, G.A.; Kanouff, M.P.; Maccallum, D.O. & Fuerschbach, P.W.
Partner: UNT Libraries Government Documents Department

Statistical Models of Mean Stress and Water Environment Effects on the Fatigue Behavior of 304 Stainless Steel

Description: Recent research efforts have focused on characterizing the effects of light water reactor environments on the fatigue behavior of austenitic stainless steels. In conjunction with these experimental programs, there has been a significant effort at Argonne National Laboratory to develop statistical models for predicting the fatigue behavior of austenitic stainless steels in air and water environments at prototypical temperatures and loading rates. Some recent testing has also been concerned with the effect of mean stress on the fatigue behavior of 304 stainless steel in air. The ultimate goal of all these efforts is to allow development of fatigue design curves and design procedures that will assure adequate margin to fatigue crack initiation under prototypical operating conditions. In this paper, a best-fit strain-life curve for 304 stainless steel in air that takes into account the effect of mean stress is developed using the Smith-Watson-Topper equivalent strain parameter. A model for predicting the effect of water environments on fatigue life in both low and high oxygen water environments for a range of temperatures and loading rates is also described. Additional effort is required to develop the most appropriate way to develop a fatigue design curve from the mean stress and water effects models.
Date: December 1, 1999
Creator: Leax, T.R.
Partner: UNT Libraries Government Documents Department

Direct Observation of Phase Transformations in Austenitic Stainless Steel Welds Using In-situ Spatially Resolved and Time-resolved X-ray Diffraction

Description: Spatially resolved x-ray diffraction (SRXRD) and time resolved x-ray diffraction (TRXRD) were used to investigate real time solid state phase transformations and solidification in AISI type 304 stainless steel gas tungsten arc (GTA) welds. These experiments were conducted at Stanford Synchrotron Radiation Laboratory (SSRL) using a high flux beam line. Spatially resolved observations of {gamma} {leftrightarrow} {delta} solid state phase transformations were performed in the heat affected zone (HAZ) of moving welds and time-resolved observations of the solidification sequence were performed in the fusion zone (FZ) of stationary welds after the arc had been terminated. Results of the moving weld experiments showed that the kinetics of the {gamma}{yields}{delta} phase transformation on heating in the HAZ were sufficiently rapid to transform a narrow region surrounding the liquid weld pool to the {delta} ferrite phase. Results of the stationary weld experiments showed, for the first time, that solidification can occur directly to the {delta} ferrite phase, which persisted as a single phase for 0.5s. Upon solidification to {delta}, the {delta} {yields} {gamma} phase transformation followed and completed in 0.2s as the weld cooled further to room temperature.
Date: September 23, 1999
Creator: Elmer, J.; Wong, J. & Ressler, T.
Partner: UNT Libraries Government Documents Department

Adsorption of Pu(IV) Polymer onto 304L Stainless Steel

Description: 'The report, Technical Basis for Safe Operations with Pu-239 Polymer in NMS S Operating Facilities (F H Areas), (WSRC-TR-99-00008) was issued in an effort to upgrade the Authorization Basis (AB) for H Area facilities relative to nuclear criticality. At the time, insufficient data were found in the literature to quantify the adsorption of Pu polymer onto the surfaces of stainless steel tanks. Additional experimental or literature information on the adsorption of Pu(IV) polymer and its removal was deemed necessary to support the H Area AB. The results obtained are also applicable to processing in F Area facilities.Additional literature sources suggest that adsorption on the tank walls should not be a safety concern. The sources show that the amount of Pu polymer that adsorbs from a solution comes to a limiting amount in 5 to 7 days after which no additional Pu is adsorbed. Adsorption increases with Pu concentration and decreases with acid concentration. The adsorbed amounts are small varying from 0.5 mg/cm2 for a 0.5 g/l Pu / 0.5M HNO3 solution to 11 mg/cm2 for a 1-3 g/l Pu / 0.1M HNO3 solution. Additionally, acid concentrations greater than 0.1M will remove a percentage of adsorbed Pu.The experimental results have generally confirmed much of what has been reported in the literature. Specifically, adsorption onto stainless steel was found to increase with increased Pu concentration, and decreased acid concentration. The amount adsorbed was found to come to a limiting amount after 5 to 7 days. Pu adsorbed as polymer was found to be harder to remove than if it was adsorbed as Pu(IV). The amount of Pu adsorbed as polymer was found to be almost an order of magnitude more than that from a similar concentration Pu(IV) solution. Unlike the literature, only a slight increase in adsorption values was found when the ...
Date: July 16, 1999
Creator: Bronikowski, M.G.
Partner: UNT Libraries Government Documents Department

External Corrosion Analysis of Model 9975 Packaging Container

Description: The Materials Consultation Group of SRTC has completed an external corrosion analysis of the Model 9975 packaging container for storage in K Reactor under ambient conditions for a period of 12 years. The 12-year storage period includes two years for shipping and ten years for storage. Based on review of existing literature and stated building storage conditions, corrosion degradation of the 304L Stainless Steel (SS) packaging container (drum and vessels) should be minimal during the 12 year time period. There may be visible corrosion on the galvanized carbon steel pallet due to initial drum handling. The visible corrosion will not be sufficient to cause significant degradation during the 12-year storage period. The Materials Consultation Group concludes that there are sufficient data to establish the technical basis for safe storage of the Model 9975 container in the 105-K building for up to 10 years following the 2-year shipping period. The data are sufficient to allow the 304L SS containers to be stored for a total period of 15 years.
Date: February 23, 1999
Creator: Vormelker, P.
Partner: UNT Libraries Government Documents Department

Containment of Nitric Acid Solutions of Plutonium-238

Description: The corrosion of various metals that could be used to contain nitric acid solutions of Pu-238 has been studied. Tantalum and tantalum/2.5% tungsten resisted the test solvent better than 304L stainless steel and several INCONEL alloys. The solvent used to imitate nitric acid solutions of Pu-238 contained 70% nitric acid, hydrofluoric acid, and ammonium hexanitratocerate.
Date: January 31, 1999
Creator: Reimus, M.A.H.; Silver, G.L.; Pansoy-Hjelvik, L. & Ramsey, K.
Partner: UNT Libraries Government Documents Department

Thermal analysis of the APT materials irradiation samples

Description: The accelerator production of tritium (APT) project proposes to use a 1.7 GeV, 100 mA proton beam to produce neutrons from an Inconel 718 clad tungsten target. The neutrons are multiplied and moderated in a lead/water blanket before being captured in He{sup 3} to form tritium. In this process, the materials in the target and blanket region are exposed to a wide range of different fluxes comprised of protons and neutrons with energies into the GeV range. To investigate the effect of irradiation on the mechanical properties of candidate APT materials (Inconel 718, 316L stainless steel, Al 6061-T6, Mod 9Cr-1Mo, 304L stainless steel and Al5052-0), the APT Engineering Design and Development group fielded an extensive materials irradiation using the LANSCE (Los Alamos Neutron Science Center) accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The test set-up was designed to place mechanical test specimens in locations in and near the proton beam where the environment of proton and neutron fluxes and temperatures are prototypic to those expected in the APT target/blanket (50--170 C). After irradiating for about 3,600 hours, the maximum achieved proton fluence was 4--5 {times} 10{sup 21}p/cm{sup 2} for the materials in the center of the beam. To obtain relevant data on the change in the mechanical properties with fluence, it is essential to know the temperature at which the materials were irradiated. This paper explains the method of determining the specimen temperature and reports some specific examples.
Date: December 31, 1998
Creator: Maloy, S.A.; Willcutt, G.J.; James, M.R.; Teague, J.; Siebe, D.A.; Sommer, W.F. et al.
Partner: UNT Libraries Government Documents Department

Development of integrated mechanistically-based degradation-mode models for performance assessment of high-level waste containers

Description: Alloy 22 [UNS NO60221] is now being considered for construction of high level waste containers to be emplaced at Yucca Mountain and elsewhere. In essence, this alloy is 20.0-22.5% Cr, 12.5-14.5% MO, 2.0-6.0% Fe, 2.5-3.5% W, with the balance being Ni. Other impurity elements include P, Si, S, Mn, Co and V. Cobalt may be present at a maximum concentration of 2.5%. Detailed mechanistic models have been developed to account for the corrosion of Alloy 22 surfaces in crevices that will inevitably form. Such occluded areas experience substantial decreases in pH, with corresponding elevations in chloride concentration. Experimental work has been undertaken to validate the crevice corrosion model, including parallel studies with 304 stainless steel.
Date: December 21, 1998
Creator: Estill, J C; Farmer, J C; Gordon, S R & McCright, R D
Partner: UNT Libraries Government Documents Department