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The PEACE PIPE: Recycling nuclear weapons into a TRU storage/shipping container

Description: This paper describes results of a contract undertaken by the National Conversion Pilot Project (NCPP) at the Rocky Flats Environmental Technology Site (RFETS) to fabricate stainless steel ``pipe`` containers for use in certification testing at Sandia National Lab, Albuquerque to qualify the container for both storage of transuranic (TRU) waste at RFETS and other DOE sites and shipping of the waste to the Waste Isolation Pilot Project (WIPP). The paper includes a description of the nearly ten-fold increase in the amount of contained plutonium enabled by the product design, the preparation and use of former nuclear weapons facilities to fabricate the components, and the rigorous quality assurance and test procedures that were employed. It also describes how stainless steel nuclear weapons components can be converted into these pipe containers, a true ``swords into plowshare`` success story.
Date: March 1, 1997
Creator: Floyd, D.; Edstrom, C.; Biddle, K.; Orlowski, R.; Geinitz, R.; Keenan, K. 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

Electrolytic decontamination of conductive materials for hazardous waste management

Description: Electrolytic removal of plutonium and americium from stainless steel and uranium surfaces has been demonstrated. Preliminary experiments were performed on the electrochemically based decontamination of type 304L stainless steel in sodium nitrate solutions to better understand the metal removal effects of varying cur-rent density, pH, and nitrate concentration parameters. Material removal rates and changes in surface morphology under these varying conditions are reported. Experimental results indicate that an electropolishing step before contamination removes surface roughness, thereby simplifying later electrolytic decontamination. Sodium nitrate based electrolytic decontamination produced the most uniform stripping of material at low to intermediate pH and at sodium nitrate concentrations of 200 g L{sup -1} and higher. Stirring was also observed to increase the uniformity of the stripping process.
Date: December 31, 1996
Creator: Wedman, D.E.; Martinez, H.E. & Nelson, T.O.
Partner: UNT Libraries Government Documents Department

Criteria determining the selection of slags for the melt decontamination of radioactively contaminated stainless steel by electroslag remelting

Description: Electroslag remelting is an excellent process choice for the melt decontamination of radioactively contaminated metals. ESR furnaces are easily enclosed and do not make use of refractories which could complicate thermochemical interactions between molten metal and slag. A variety of cleaning mechanisms are active during melting; radionuclides may be partitioned to the slag by means of thermochemical reaction, electrochemical reaction, or mechanical entrapment. At the completion of melting, the slag is removed from the furnace in solid form. The electroslag process as a whole is greatly affected by the chemical and physical properties of the slag used. When used as a melt decontamination scheme, the ESR process may be optimized by selection of the slag. In this research, stainless steel bars were coated with non-radioactive surrogate elements in order to simulate surface contamination. These bars were electroslag remelted using slags of various chemistries. The slags investigated were ternary mixtures of calcium fluoride, calcium oxide, and alumina. The final chemistries of the stainless steel ingots were compared with those predicted by the use of a Free Energy Minimization Modeling technique. Modeling also provided insight into the chemical mechanisms by which certain elements are captured by a slag. Slag selection was also shown to have an impact on the electrical efficiency of the process as well as the surface quality of the ingots produced.
Date: March 1, 1997
Creator: Buckentin, J.M.R.; Damkroger, B.K.; Shelmidine, G.J. & Atteridge, D.G.
Partner: UNT Libraries Government Documents Department

Analysis of cracking of co-extruded recovery boiler floor tubes

Description: Cracking of the stainless steel layer in co-extruded 304L/SA210 tubing used in black liquor recovery boilers is being found in an ever-increasing number of North American pulp and paper mills. Because of the possibility of a tube failure, this is a significant safety issue, and, because of the extra time required for tube inspection and repair, this can become an economic issue as well. In a project funded by the U.S. Department of Energy and given wide support among paper companies, boiler manufacturers, and tube fabricators, studies are being conducted to determine the cause of the cracking and to identify alternate materials and/or operating procedures to prevent tube cracking. Examination of cracked tubes has permitted characterization of crack features, and transmission electron microscopy is providing information about the thermal history, particularly cyclic thermal exposures, that tubes have experienced. Neutron and x-ray diffraction techniques are being used to determine the residual stresses in as-fabricated tube panels and exposed tubes, and finite element modeling is providing information about the stresses the tubes experience during operation. Laboratory studies are being conducted to determine the susceptibility of the co-extruded 304L/SA210 tubes to stress corrosion cracking, thermal fatigue, and corrosion in molten smelt. This paper presents the current status of these studies. On the basis of all of these studies, recommendations for means to prevent tube cracking will be offered.
Date: August 1, 1997
Creator: Keiser, J.R.; Taljat, B. & Wang, X.L.
Partner: UNT Libraries Government Documents Department

Spallation source materials test program

Description: A spallation source materials program has been developed to irradiate and test candidate materials (Inconel 718, 316L and 304L stainless steel, modified 9Cr-1Mo(T91), Al6061-T6, Al5052-O) for use in the Accelerator Production of Tritium (APT) target and blanket in prototypic proton and neutron fluxes at prototypic temperatures. The study uses the 800 MeV, 1mA proton accelerator at the Los Alamos Neutron Science Center (LANSCE) which produces a Gaussian beam with 2 sigma = 3 cm. The experimental set-up contains prototypic modules of the tungsten neutron source and the lead/aluminum blanket with mechanical testing specimens of candidate APT materials placed in specific locations in the irradiation area. These specimens have been irradiated for greater than 3,600 hours with a maximum proton fluence of 4--5 {times} 10{sup 21} p/cm{sup 2} in the center of the proton beam. Specimens will yield some of the first data on the effect of proton irradiation to high dose on the materials` properties from tensile tests, 3 pt. bend tests, fracture toughness tests, pressurized tubes, U-bend stress corrosion cracking specimens, corrosion measurements and microstructural characterization of transmission electron microscopy specimens.
Date: December 1, 1997
Creator: Maloy, S.A. & Sommer, W.F.
Partner: UNT Libraries Government Documents Department

Application of internal state variable plasticity and damage models to welding

Description: An internal state variable constitutive model coupled with a ductile void growth model is applied in two finite element simulations of welding. Shrinkage of a 304 L stainless steel pipe due to multipass gas tungsten arc welds is presented as an example of tracking distortion far from the weld. Weld solidification cracking in Al-6061 disks is presented as an application of the plasticity model coupled with the damage model.
Date: June 1, 1997
Creator: Dike, J.J.; Ortega, A.R.; Bammann, D.J. & Lathrop, J.F.
Partner: UNT Libraries Government Documents Department

Impact of phase stability on the corrosion behavior of the austenitic candidate materials for NNWSI [Nevada Nuclear Waste Storage Investigations]

Description: The Nuclear Waste Management Program at Lawrence Livermore National Laboratory is responsible for the development of the waste package design to meet the Nuclear Regulatory Commission licensing requirements for the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. The metallic container component of the waste package is required to assist in providing substantially complete containment of the waste for a period of up to 1000 years. Long term phase stability of the austenitic candidate materials (304L and 316L stainless steels and alloy 825) over this time period at moderate temperatures (100-250{sup 0}C) can impact the mechanical and corrosion behavior of the metal barrier. A review of the technical literature with respect to phase stability of 304L, 316L and 825 is presented. The impact of martensitic transformations, carbide precipitation and intermediate ({sigma}, chi, and eta) phase formation on the mechanical properties and corrosion behavior of these alloys at repository relevant conditions is discussed. The effect of sensitization on intergranular stress corrosion cracking (IGSCC) of each alloy is also addressed. A summary of the impact of phase stability on the degradation of each alloy in the proposed repository environment is included. 32 refs., 6 figs.
Date: October 1, 1987
Creator: Bullen, D.B.; Gdowski, G.E. & McCright, R.D.
Partner: UNT Libraries Government Documents Department

Sensors for feedback controls in solid state resistance welding operations

Description: The Savannah River Site (SRS) has a 40-plus year history of producing and processing tritium primarily for use in nuclear weapons. This gas is stored at high pressures in reservoirs that are manufactured and sealed through the use of special resistance welding processes. There is an interest in maintaining the quality and consistency of these welds to avoid leaks in the reservoirs. The reasons for this are the limited supply and high cost of producing tritium, the necessity of assuring nuclear safety and to promote weapon system reliability. Precisely machined 304-L and 316 stainless steel components are the materials used in the fabrication of the reservoir. This presentation will include a survey of sensors for use in resistance welding processes. The results of the application of the analog laser position sensor will be presented along with data indicating how the displacement parameter defines the weld process. Opportunities to close the control loop by taking sensor data into the weld controller will be discussed.
Date: October 1, 1997
Creator: Tarpley, J.
Partner: UNT Libraries Government Documents Department

Causes and solutions for cracking of coextruded and weld overlay floor tubes in black liquor recovery boilers

Description: Cracking of coextruded, black liquor recovery boiler floor tubes is both a safety and an economic issue to mill operators. In an effort to determine the cause of the cracking and to identify a solution, extensive studies, described in this and three accompanying papers, are being conducted. In this paper, results of studies to characterize both the cracking and the chemical and thermal environment are reported. Based on the results described in this series of papers, a possible mechanism is presented and means to lessen the likelihood of cracking or to totally avoid cracking of floor tubes are offered.
Date: September 1, 1998
Creator: Keiser, J.R.; Taljat, B. & Wang, X.L.
Partner: UNT Libraries Government Documents Department

Finite Element Modeling and Validation of Residual Stresses in 304L Girth Welds

Description: Three dimensional finite element simulations of thermal and mechanical response of a 304L stainless steel pipe subjected to a circumferential autogenous gas tungsten arc weld were used to predict residual stresses in the pipe. Energy is input into the thermal model using a volumetric heat source. Temperature histories from the thermal analysis are used as loads in the mechanical analyses. In the mechanical analyses, a state variable constitutive model was used to describe the material behavior. The model accounts for strain rate, temperature, and load path histories. The predicted stresses are compared with X-ray diffraction determinations of residual stress in the hoop and axial directions on the outside surface of the pipe. Calculated stress profiles fell within the measured data. Reasons for the observed scatter in the measured stresses are discussed.
Date: June 1, 1998
Creator: Dike, J. J.; Ortega, A. R. & Cadden, C. H.
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

Material Selection for Defense Waste Processing Facility

Description: Construction has started on a facility to immobilize high-level radioactive waste in borosilicate glass at the Department of Energy's Savannah River Plant. Type 304L stainless steel is generally sufficient for supply tankage and service lines. It is used as the reference material in chemical reprocessing of reactor target and fuel tubes. Type 304L, however, has unacceptable stress corrosion cracking resistance in solutions containing formic acid and chloride. Scouting tests were performed on twelve commercial nickel-based alloys in simulated process solutions containing halides, sulfates, nitrates, mercury and formic acid. Mercuric ions and halides interact in acidic environments to increase pitting and crevice attack. Alloys with combined chromium plus molybdenum contents greater than 30 percent, that also contain greater than 9 percent molybdenum, were most resistant to pitting and crevice corrosion. Based on this testing, Alloy C-276 has been selected as the reference process equipment material, with Inconel 690 and ALLCORR selected for specialty areas.
Date: July 17, 1985
Creator: Bickford, D.F.
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

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

The effect of travel speed on thermal response in CO{sub 2} laser welding of small electronic components

Description: A comprehensive three-dimensional numerical investigation of the effect of beat source travel speed on temperatures and resulting thermal stresses was performed for CO{sub 2}-laser welding. The test specimen was a small thermal battery header containing several stress-sensitive glass-to-metal seals surrounding the electrical connections and a temperature sensitive ignitor located under the header near the center. Predictions of the thermal stresses and temperatures in the battery header were made for several travel speeds of the laser. The travel speeds examined ranged from 10mm/sec to 50mm/sec. The results indicate that faster weld speeds result in lower temperatures and stresses for the same size weld. This is because the higher speed welds are more efficient, requiring less energy to produce a given weld. Less energy absorbed by the workpiece results in lower temperatures, which results in lower stresses.
Date: February 1, 1995
Creator: Gianoulakis, S.E.; Burchett, S.N.; Fuerschbach, P.W. & Knorovsky, G.A.
Partner: UNT Libraries Government Documents Department

Selection of barrier metals for a waste package in tuff

Description: The Nevada Nuclear Waste Storage Investigation (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. LLNL is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. The selection of metal containment barriers is addressed. 13 references.
Date: September 1, 1983
Creator: Russell, E.W.; McCright, R.D. & O`Neal, W.C.
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

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

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

Microstructural and solidification cracking evaluation of electron beam welds in 304L

Description: Weld hot cracking of stainless steels is a major materials-related problem in the welding industry. This present investigation evaluates the crack susceptibility of highly-constrained EB welds made in materials whose DeLong ferrite potentials range from zero to nine FN. In addition, the effect of piece part strength level on cracking is examined. This study has revealed that these deep penetration EB welds have regions that solidify as primary austenite, even when the DeLong ferrite potential is as high as 9 FN. This points out the critical role that solidification rate plays in the crack susceptibility of these highly restrained welds. In addition, 0 FN to 0 FN welds had primarily transverse cracks while 6 FN to 0 FN welds had primarily centerline cracks. Of particular interest is the observation that cracks still occur if a high ferrite (greater than 6 FN) component is welded to a zero FN component. Cracking is always associated with regions which solidify as primary austenite and these cracks occur because there are areas in the weld which do not mix. Thus it is not a recommended production practice to compensate for low ferrite in one piece part with high ferrite in its mate. Finally, it is shown that a DeLong FN threshold of 4 to prevent cracking in EB welds in not valid. 21 refs., 16 figs.
Date: January 1, 1991
Creator: Sturgill, P.L.; Campbell, R.D. & Henningsen, J.L.
Partner: UNT Libraries Government Documents Department

J-controlled crack growth as an indicator of hydrogen-stainless steel compatibility

Description: The J-integral was evaluated as a parameter to characterize fracture of stainless steels and as a measure of hydrogen damage. C-shaped specimens of type 304L, 316, and 21-6-9 stainless steels were tested in high pressure helium and hydrogen. The critical force for crack initiation (Jm), and tearing resistance (dJ/da) were decreased by hydrogen in all three alloys. The J-integral appears useful as a measure of hydrogen compatibility because it is sensitive to both test environment and microstructure.
Date: January 1, 1980
Creator: Dietrich, M.R.; Caskey, G.R. Jr. & Donovan, J.A.
Partner: UNT Libraries Government Documents Department