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Potentiodynamic polarization studies on candidate container alloys for the Tuff Repository

Description: Cortest Columbus Technologies, Inc. (CC Technologies) is investigating the long-term performance of container materials used for high-level radioactive waste packages. This information is being developed for the Nuclear Regulatory Commission to aid in their assessment of the Department of Energy`s application to construct a geologic repository for disposal of high-level radioactive waste. This report summarizes the results of cyclic-potentiodynamic-polarization (CCP) studies performed on candidate container materials for the Tuff Repository. The CPP technique was used to provide an understanding of how specific variables such as environmental composition, temperature, alloy composition, and welding affect both the general- and localized-corrosion behavior of two copper-base and two Fe-Cr-Ni alloys in simulated repository environments. A statistically-designed test solution matrix was formulated, based on an extensive search of the literature, to evaluate the possible range of environmental species that may occur in the repository over the life of the canister. Forty-two CPP curves were performed with each alloy and the results indicated that several different types of corrosion were possible. The copper-base alloys exhibited unusual CCP behavior in that hysteresis was not always associated with pitting. The effects of temperature on the corrosions behavior were evaluated in two types of tests; isothermal tests at temperatures from 50{degrees}C to 90{degrees}C and heat-transfer tests where the solution was maintained at 50{degrees}C and the specimen was internally heated to 90{degrees}C. In the isothermal test, CPP curves were obtained with each alloy in simulated environments at 50{degrees}C, 75{degrees}C, and 90{degrees}C. The results of these CCP experiments indicated that no systematic trends were evident for the environments tested. Lastly, the effects of welding on the corrosion behavior of the alloys in simulated environments were examined.
Date: January 1, 1992
Creator: Thompson, N.G.; Beavers, J.A. & Durr, C.L.
Partner: UNT Libraries Government Documents Department

Stress corrosion cracking of candidate waste container materials

Description: Six alloys have been selected as candidate container materials for the storage of high-level nuclear waste at the proposed Yucca Mountain site in Nevada. These materials are Type 304L stainless steel (SS), Type 316L SS, Incology 825, P-deoxidized Cu, Cu-30%Ni, and Cu-7% Al. The present program has been initiated to determine whether any of these materials can survive for 300 years in the site environment without developing through-wall stress corrosion cracks, and to assess the relative resistance of these materials to stress corrosion cracking (SCC). A series of slow-strain-rate tests (SSRTs) in simulated Well J-13 water which is representative of the groundwater present at the Yucca Mountain site has been completed, and crack-growth-rate (CGR) tests are also being conducted under the same environmental conditions. 13 refs., 60 figs., 22 tabs.
Date: November 1, 1990
Creator: Maiya, P.S.; Soppet, W.K.; Park, J.Y.; Kassner, T.F.; Shack, W.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Localized weld metal corrosion in stainless steel water tanks

Description: The rapidly developed leaks within the TFC and TFD tanks (LLNL groundwater treatment facilities) were caused by localized corrosion within the resolidified weld metal. The corrosion was initiated by the severe oxidation of the backsides of the welds which left the exposed surfaces in a condition highly susceptible to aqueous corrosion. The propagation of surface corrosion through the thickness of the welds occurred by localized corrosive attack. This localized attack was promoted by the presence of shielded aqueous environments provided by crevices at the root of the partial penetration welds. In addition to rapid corrosion of oxidized surfaces, calcium carbonate precipitation provided an additional source of physical shielding from the bulk tank environment. Qualification testing of alternate weld procedures showed that corrosion damage can be prevented in 304L stainless steel GTA welds by welding from both sides while preventing oxidation of the tank interior through the use of an inert backing gas such as argon. Corrosion resistance was also satisfactory in GMA welds in which oxidized surfaces were postweld cleaned by wire brushing and chemically passivated in nitric acid. Further improvements in corrosion resistance are expected from a Mo-containing grade of stainless steel such as type 316L, although test results were similar for type 304L sheet welded with type 308L filler metal and type 316L sheet welded with type 316L filler metal.
Date: May 25, 1995
Creator: Strum, M.J.
Partner: UNT Libraries Government Documents Department

Crack-growth-rate testing of candidate waste container materials

Description: Fracture-mechanics crack growth tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L stainless steel (SS) and Incoloy 825 at 93{degree}C and 1 atmosphere of pressure is simulated J-13 well water, which is representative of the groundwater at the Yucca Mountain site in Nevada that is proposed for a high-level nuclear waste repository. Crack growth rates were measured under various load conditions: load ratios (R) of 0.5--1.0, frequencies of 10{sup {minus}3}{minus}1 Hz, rise times of 1--1000 s, and peak stress intensities of 25--40 MPa{center_dot}m{sup 1/2}. The measured crack growth rates are bounded by the predicted rates from the current ASME Section XI correlation for fatigue crack growth rates of austenitic stainless steel in air. Environmentally accelerated crack growth was not evident in any of the three materials under the test conditions investigated.
Date: December 31, 1991
Creator: Park, J.Y.; Shack, W.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Stress corrosion cracking of candidate waste container materials; Final report

Description: Six alloys have been selected as candidate container materials for the storage of high-level nuclear waste at the proposed Yucca mountain site in Nevada. These materials are Type 304L stainless steel (SS). Type 316L SS, Incoloy 825, phosphorus-deoxidized Cu, Cu-30%Ni, and Cu-7%Al. The present program has been initiated to determine whether any of these materials can survive for 300 years in the site environment without developing through-wall stress corrosion cracks. and to assess the relative resistance of these materials to stress corrosion cracking (SCC)- A series of slow-strain-rate tests (SSRTs) and fracture-mechanics crack-growth-rate (CGR) tests was performed at 93{degree}C and 1 atm of pressure in simulated J-13 well water. This water is representative, prior to the widespread availability of unsaturated-zone water, of the groundwater present at the Yucca Mountain site. Slow-strain-rate tests were conducted on 6.35-mm-diameter cylindrical specimens at strain rates of 10-{sup {minus}7} and 10{sup {minus}8} s{sup {minus}1} under crevice and noncrevice conditions. All tests were interrupted after nominal elongation strain of 1--4%. Scanning electron microscopy revealed some crack initiation in virtually all the materials, as well as weldments made from these materials. A stress- or strain-ratio cracking index ranks these materials, in order of increasing resistance to SCC, as follows: Type 304 SS < Type 316L SS < Incoloy 825 < Cu-30%Ni < Cu and Cu-7%Al. Fracture-mechanics CGR tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L stainless steel (SS) and Incoloy 825. Crack-growth rates were measured under various load conditions: load ratios M of 0.5--1.0, frequencies of 10{sup {minus}3}-1 Hz, rise nines of 1--1000s, and peak stress intensities of 25--40 MPa{center_dot}m {sup l/2}.
Date: June 1, 1992
Creator: Park, J.Y.; Maiya, P.S.; Soppet, W.K.; Diercks, D.R.; Shack, W.J. & Kassner, T.F.
Partner: UNT Libraries Government Documents Department

Application of the NNWSI [Nevada Nuclear Waste Storage Investigations] unsaturated test method to actinide doped SRL [Savannah River Laboratory] 165 type glass

Description: The results of tests done using the Unsaturated Test Method are presented. These tests, done to determine the suitability of glass in a potential high-level waste repository as developed by the Nevada Nuclear Waste Storage Investigations Project, simulate conditions anticipated for the post-containment phase of the repository when only limited contact between the waste form and water is expected. The reaction of glass occurs via processes that are initiated due to glass/water vapor and glass/liquid water contact. Vapor interaction results in the initiation of an exchange process between water and the more mobile species (alkalis and boron) in the glass. The liquid reaction produces interactions similar to those seen in standard leaching tests, except due to the limited amount of water present and the presence of partially sensitized 304L stainless steel, the formation of reaction products greatly exceeds that found in MCC-1 type leach tests. The effect of sensitized stainless steel on the reaction is to enhance breakdown of the glass matrix thereby increasing the release of the transuranic elements from the glass. However, most of the Pu and Am released is entrained by either the metal components of the test or by the reaction phases, and is not released to solution. 16 refs., 20 figs., 17 tabs.
Date: August 1, 1990
Creator: Bates, J.K. & Gerding, T.J.
Partner: UNT Libraries Government Documents Department

The effects of hydrogen on the fracture toughness properties of upset welded stainless steel

Description: The effects of hydrogen on the fracture toughness properties of upset welded Type 304L stainless steel were measured and compared to those measured previously for as-received and as-welded steels. The results showed that the upset welded steels had good fracture toughness properties, but values were lower than the as-received material. The fracture toughness value of the base material was 6420 in-lbs/sq. in., while the welded steels averaged 3660 in-lbs/sq. in. Hydrogen exposure lowered the fracture toughness values of the as-received steel by 43 % to 3670 in-lbs/sq. in. and the welded steels by 21 % to 2890 in-lbs/sq. in. The fracture morphologies of the unexposed steels showed that ductile fracture occurred by the microvoid nucleation and growth process. The size of the microvoids on the fracture surfaces of the welded steels were much smaller and more closely spaced that those found on the base material fracture surfaces. The change in the size and spacing of the microvoids indicates that the fracture toughness properties of the welded steels were lower than the base steels because of the higher concentration of microscopic precipitates on the weld plane. The welds examined thus far have been {open_quotes}good{close_quotes} welds and the presence of these precipitates was not apparent in standard {open_quotes}low{close_quotes}-magnification metallographic sections of the weld planes. The results indicate that hydrogen did not weaken greatly the solid-state welds but that other inclusions or impurities present prior to welding did. Improvements in surface cleaning and preparation prior to welding should be explored as a way to improve the strength of solid-state welded joints.
Date: June 1, 1995
Creator: Morgan, M.T.
Partner: UNT Libraries Government Documents Department

The reaction of SRL 202 glass in J-13 and DIW

Description: Static leach tests were performed in both 304L stainless steel and Teflon vessels using a synthetic high-level waste glass with either deionized water (DIW) or a tuff groundwater solution as the leachant to assess the effects of the vessel and the initial leachant composition on the extent and nature of the glass reaction. The tests were performed using monolith samples at 340 m{sup {minus}1} and crushed samplesat 2000 m{sup {minus}1} for times up to 1 year. The results show less silicon is released from the glass into the groundwater solution than into DIW at both high and low glass surface area/leachant volume ratios (SAN), but the alkali metal and boron releases are not affected by the leachant used. Tests performed in a stainless steel vessel resulted in slightly lower leachate pH values, but similar reaction rates to those performed in a Teflon vessel, as measured by the boron release. Blank tests with DIW or EJ-13 in the vessels showed the Teflon vessels to release small amounts of fluoride (1 to 2 ppm) and to acidify the DIW slightly (4.0 < pH < 5.6). The pH values of blank tests with EJ- 1 3 increased from 8.2 to about 8.6 in steel and to about 9.2 in Teflon vessels. The slightly higher pH values attained in Teflon vessels are attributed to outgassing of CO{sub 2} during the test.
Date: December 31, 1992
Creator: Ebert, W.L.; Bates, J.K. & Buck, E.C.
Partner: UNT Libraries Government Documents Department

Yucca Mountain project container fabrication, closure and non-destructive evaluation development activities; Summary and viewgraphs

Description: In this presentation, container fabrication, closure, and non-destructive evaluation (NDE) process development activities are described. All of these activities are interrelated, and will contribute to the metal barrier selection activity. The plan is to use a corrosion-resistant material in the form of a cylinder with a wall thickness of {approximately}1cm (2cm for pure copper.) The materials under consideration include the three austenitic alloys: stainless steel-304L, stainless steel-316L and alloy 825, as well as the three copper alloys: CDA 102, CDA 613, and CDA 715. This document reviews the recommended procedures and processes for fabricating, closing and evaluating each of the candidate materials. (KGD)
Date: June 1989
Creator: Russell, E. W. & Nelson, T. A.
Partner: UNT Libraries Government Documents Department

A technical basis to relax the dew point specification for the environment in the vapor space in DWPF canisters

Description: This memorandum establishes the technical basis to conclude that relaxing, from 0 C to 20 C, the dew point specification for the atmosphere in the vapor space (free volume) of a DWPF canister will not provide an environment that will cause significant amounts of corrosion induced degradation of the canister wall. The conclusion is based on engineering analysis, experience and review of the corrosion literature. The basic assumptions underlying the conclusion are: (1) the canister was fabricated from Type 304L stainless steel; (2) the corrosion behavior of the canister material, including base metal, fusion zones and heat effected zones, is typified by literature data for, and industrial experience with, 300 series austenitic stainless steels; and (3) the glass-metal crevices created during the pouring operation will not alter the basic corrosion resistance of the steel although such crevices might serve as sites for the initiation of minor amounts of corrosion on the canister wall.
Date: May 1, 1995
Creator: Louthan, M.R. Jr.
Partner: UNT Libraries Government Documents Department

Crack growth behavior of candidate waste container materials in simulated underground water

Description: Fracture-mechanics crack growth tests were conducted on 25.4-mm-thick compact tension specimens of Types 304L and 316L Stainless steel and Incoloy 825 at 93{degrees}C and 1 atmosphere of pressure in simulated J-13 well water, which is representative of the groundwater at the Yucca Mountain site in Nevada that is proposed for a high-level nuclear waste repository. Crack growth rates were measured under various load conditions: load ratios of 0.2--1.0, frequencies of 2 {times} 10{sup {minus}4}{minus}1 Hz, rise times of 1--5000 s, and peak stress intensities of 25--40 MPa{center_dot}m{sup {1/2}}. The measured crack growthrates are bounded by the predicted rates from the current ASME Section 11 correlation for fatigue crack growth rates of austenitic stainless steel in air. Environmentally accelerated crack growth was not evident in any of the three materials under the test conditions investigated.
Date: December 31, 1992
Creator: Park, J.Y.; Shack, W.J. & Diercks, D.R.
Partner: UNT Libraries Government Documents Department

Parametric effects of glass reaction under unsaturated conditions

Description: Eventual liquid water contact of high-level waste glass stored under the unsaturated conditions anticipated at the Yucca Mountain site will be by slow intrusion of water into a breached container/canister assembly. The water flow patterns under these unsaturated conditions will vary, and the Unsaturated Test method has been developed by the YMP to study glass reaction. The results from seven different sets of tests done to investigate the effect of systematically varying parameters, such as glass composition, composition and degree of sensitization of 304L stainless steel, water input volume, and the interval of water contact are discussed. Glass reaction has been monitored over a period of five years, and the parametric effects can result in up to a ten-fold variance in the degree of glass reaction.
Date: November 1, 1989
Creator: Bates, J.K.; Gerding, T.J. & Woodland, A.B.
Partner: UNT Libraries Government Documents Department

Stress-corrosion-cracking studies on candidate container alloys for the Tuff Repository

Description: Cortest Columbus Technologies, Inc. (CC Technologies) investigated the long-term performance of container materials used for high-level waste package as part of the information needed by the Nuclear Regulatory Commission (NRC) to assess the Department of Energy`s application to construct to geologic repository for high-level radioactive waste. At the direction of the NRC, the program focused on the Tuff Repository. This report summarizes the results of Stress-Corrosion-Cracking (SCC) studies performed in Tasks 3, 5, and 7 of the program. Two test techniques were used; U-bend exposures and Slow-Strain-Rate (SSR) tests. The testing was performed on two copper-base alloys (Alloy CDA 102 and Alloy CDA 175) and two Fe-Cr-Ni alloys (Alloy 304L and Alloy 825) in simulated J-13 groundwater and other simulated solutions for the Tuff Repository. These solutions were designed to simulate the effects of concentration and irradiation on the groundwater composition. All SCC testing on the Fe-Cr-Ni Alloys was performed on solution-annealed specimens and thus issues such as the effect of sensitization on SCC were not addressed.
Date: May 1, 1992
Creator: Beavers, J.A. & Durr, C.L.
Partner: UNT Libraries Government Documents Department

Prevention for possible microbiologically influenced corrosion (MIC) in RHLWE flush water system

Description: This report is in response to the request to provide a recommendation for the prevention of possible microbiologically influenced corrosion (MIC) for the RHLWE (Replacement High-Level Waste Evaporator) flush water (FW) system. The recent occurrences of MIC at DWPF prompted HLWE to evaluate the possibility of MIC occurring in this 304L stainless steel RHLWE flush water system. Concern was heightened by the fact that the well water used and the other conditions at H-Tank Farm are similar to those at DWPF. However, only one known leak has occurred in the existing 304L evaporator flush water systems in either tank farm (in 1H system), and no MIC Corrosion has been confirmed in the tank farm area. The design of the RHLWE flush water system (completed long before the occurrence of MIC at DWPF) was modeled after the existing evaporator flush water systems and did not specifically include MIC prevention considerations. Therefore, MIC prevention was not specifically considered during the design phase of this flush water system. The system is presently being installed. After an extensive evaluation, a task team concluded that the best biocide to prevent the occurrence of MIC would be NaOH at fairly low concentration. Sodium hydroxide (NaOH) is optimal in this application, because of its effectiveness, low cost, and familiarity to the Operations personnel (see Appendix A). However, it is the opinion of the task group that application should be withheld until MIC corrosion is demonstrated in the system.
Date: July 10, 1995
Creator: Hsu, T.C. & Jenkins, C.F.
Partner: UNT Libraries Government Documents Department

Fabrication and closure development of nuclear waste disposal containers for the Yucca Mountain Project: Status report

Description: In GFY 89, a project was underway to determine and demonstrate a suitable method for fabricating thin-walled monolithic waste containers for service within the potential repository at Yucca Mountain. A concurrent project was underway to determine and demonstrate a suitable closure process for these containers after they have been filled with high-level nuclear waste. Phase 1 for both the fabrication and closure projects was a screening phase in which candidate processes were selected for further laboratory testing in Phase 2. This report describes the final results of the Phase 1 efforts. It also describes the preliminary results of Phase 2 efforts.
Date: September 1, 1991
Creator: Domian, H.A.; Robitz, E.S.; Conrardy, C.C.; LaCount, D.F.; McAninch, M.D.; Fish, R.L. et al.
Partner: UNT Libraries Government Documents Department

A transmission electron microscopy evaluation of solid-state upset welds in Type 304L stainless steel

Description: Transmission electron microscopy (TEM) was used to characterize the microstructures at and near the weld interface in upset welded Type 304L stainless steel test samples. Two sample configurations were examined in this study; upset welded cylinders prepared using a commercial resistance welder and cylindrical shaped samples welded in a Gleeble 1500 thermomechanical simulation device. The Gleeble samples evaluated were welded at 800 C, 900 C and 1,200 C with a 0.5 cm weld upset. The base microstructure of the samples varied with weld temperature. The lower temperature specimens contained a large free-dislocation density and distinct dislocation cells. The higher temperature specimens contained well-developed subgrains and a much lower free-dislocation density. The microstructure of the upset welded samples most closely resembled the 1,200 C Gleeble sample. No distinct bond line was observed by TEM in any of the specimens, i.e., diffusion and grain growth occurred across all weld interfaces. However, weld interfaces in both specimen configurations were characterized by the presence of 50--300 nm diameter particles spaced between 300 and 1,300 nm apart. Through the use of electron diffraction analysis and X-ray microanalysis two precipitate types were identified in both specimen configurations. A crystalline phase very similar to Mn{sub 1.5}Cr{sub 1.5}O{sub 4} and an amorphous phase enriched mainly in Si and Al were observed. Surface oxides and/or internal impurities may be sources for these precipitates. Future work will include a controlled study designed to determine the origin of the interface precipitates.
Date: September 8, 1995
Creator: Tosten, M.H.
Partner: UNT Libraries Government Documents Department

Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud

Description: Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder.
Date: June 1, 1996
Creator: Payzant, A.; Spooner, S.; Zhu, Xiaojing & Hubbard, C.R.
Partner: UNT Libraries Government Documents Department

Packaging radioactive wastes for geologic disposal

Description: The M&O contractor for the DOE Office of Civilian Radioactive Waste Management is developing designs of waste packages that will contain the spent nuclear fuel assemblies from commercial and Navy reactor plants and various civilian and government research reactor plants, as well as high-level wastes vitrified in glass. The safe and cost effective disposal of the large and growing stockpile of nuclear waste is of national concern and has generated political and technical debate. This paper addresses the technical aspects of disposing of these wastes in large and robust waste packages. The paper discusses the evolution of waste package design and describes the current concepts. In addition, the engineering and regulatory issues that have governed the development are summarized and the expected performance in meeting the requirements are discussed.
Date: August 1, 1996
Creator: Benton, H.A.
Partner: UNT Libraries Government Documents Department

Sealing 304L to lithia-alumina-silica (LAS) glass-ceramics

Description: The formation of a crack-free between 300 series stainless steel and a glass-ceramic is difficult owing to the high coefficients of thermal expansion of the stainless steels. Lithia-alumina-silica (LAS) glass-ceramics were successfully developed and sealed to 304L stainless steel. These crack-free seals were fabricated by two techniques: by adjusting the parent glass composition (reducing the Al{sub 2}O{sub 3} content), or by adjusting the sealing/crystallization cycle. All seals were hermetic, with leak rates < 10{sup -8} cc/sec STP helium. CTE and alloy yield strengths are given which show the feasibility of using these materials to make feedthroughs, pyrotechnic components, etc. Metallography, SEM, and wavelength dispersive spectroscopy show the quality and integrity of the glass-ceramic/stainless steel interface. These data are compared to those on the Inconel 718/LAS-glass seal system.
Date: December 31, 1989
Creator: Moddeman, W.E.; Pence, R.E.; Massey, R.T.; Cassidy, R.T. & Kramer, D.P.
Partner: UNT Libraries Government Documents Department

Scenarios for the Evaluation of the Criticality Potential of High Actinide Glasses

Description: Vitrification is one of the leading options for immobilization of actinide-containing materials no longer needed for national defense. For these glasses to be suitable for disposal, it must be established that a significant potential for a nuclear criticality involving these glasses does not exist. The vitrification working group within the nuclear materials disposition program has been given the responsibility for developing scenarios to be evaluated. In this report, potential bounding scenarios for disposal of high actinide glasses in a geologic setting are described. These scenarios are being provided to the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM) so that the potential for criticality can be evaluated. If the evaluation of these scenarios by OCRWM reveals a significant potential for criticality then a sensitivity analysis to numerical values should be used to determine whether more precise definitions of any parameter is warranted. It is anticipated that there will need to be extensive interactions between the working group and the personnel performing the criticality evaluations.
Date: January 19, 1996
Creator: Plodinec, M.J.
Partner: UNT Libraries Government Documents Department

Upset welded 304L and 316L vessels for storage tests

Description: Two sets of vessels for tritium storage tests were fabricated using upset welding. A solid-state resistance upset weld was used to join the two halves of each vessel at the girth. The vessels differ from production reservoirs in design, material, and fabrication process. One set was made from forged 304L stainless steel and the other from forged 316L stainless steel. Six vessels of each type were loaded with a tritium mix in November 1995 and placed in storage at 71 C. This memo describes and documents the fabrication of the twelve vessels.
Date: April 1996
Creator: Kanne, W. R., Jr.
Partner: UNT Libraries Government Documents Department

Analysis of composite tube cracking in recovery boiler floors

Description: Cracking of co-extruded (generally identified as composite) floor tubes in kraft black liquor recovery boilers was first observed in Scandinavia, but this problem has now been found in many North American boilers. In most cases, cracking in the outer 304L stainless steel has not progressed into the carbon steel, but the potential for such crack propagation is a cause of concern. A multidimensional study has been initiated to characterize the cracking seen in composite floor tubes, to measure the residual stresses resulting from composite tube fabrication, and to predict the stresses in tubes under operating conditions. The characterization studies include review of available reports and documents on composite tube cracking, metallographic examination of a substantial number of cracked tubes, and evaluation of the dislocation structure in cracked tubes. Neutron and X-ray diffraction are being used to determine the residual stresses in composite tubes from two major manufacturers, and finite element analysis is being used to predict the stresses in the tubes during normal operation and under conditions where thermal fluctuations occur.
Date: August 1, 1996
Creator: Keiser, J.R.; Taljat, B.; Wang, X.L.; Maziasz, P.J.; Hubbard, C.R.; Swindeman, R.W. et al.
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

Materials property testing using a stress-strain microprobe

Description: The Stress-Strain Microprobe (SSM) uses an automated ball indentation technique to obtain flow data from a localized region of a test specimen or component. This technique is used to rapidly determine the yield strength and microstructural condition of a variety of materials including pressure vessel steels, stainless steels, and nickel-base alloys. The SSM provides an essentially non-destructive technique for the measurement of yield strength data. This technique is especially suitable for the study of complex or highly variable microstructures such as weldments and weld heat affected zones. In this study 119 distinct SSM determinations of the yield strength of eight engineering alloys are discussed and compared to data obtained by conventional tensile tests. The sensitivity of the SSM to the presence of residual stresses is also discussed.
Date: September 1, 1998
Creator: Panayotou, N.F.; Baldrey, D.G. & Haggag, F.M.
Partner: UNT Libraries Government Documents Department