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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

Analysis of components from drip tests with ATM-10 glass

Description: Waste package assemblies consisting of actinide-doped West Valley ATM-10 reference glass and sensitized 304L stainless steel have been reacted with simulated repository groundwater using the Unsaturated Test Method. Analyses of surface corrosion and reaction products resulting from tests that were terminated at scheduled intervals between 13 and 52 weeks are reported. Analyses reveal complex interactions between the groundwater, the sensitized stainless steel waste form holder, and the glass. Alteration phases form that consist mainly of smectite clay, brockite, and an amorphous thorium iron titanium silicate, the latter two incorporating thorium, uranium, and possibly transuranics. The results from the terminated tests, combined with data from tests that are still ongoing, will help determine the suitability of glass waste forms in the proposed high-level repository at the Yucca Mountain Site.
Date: September 1, 1996
Creator: Fortner, J.A.; Bates, J.K. & Gerding, T.J.
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

Initial specifications for nuclear waste package external dimensions and materials

Description: Initial specifications of external dimensions and materials for waste package conceptual designs are given for Defense High Level Waste (DHLW), Commercial High Level Waste (CHLW) and Spent Fuel (SF). The designs have been developed for use in a high-level waste repository sited in a tuff media in the unsaturated zone. Drawings for reference and alternative package conceptual designs are presented for each waste form for both vertical and horizontal emplacement configurations. Four metal alloys: 304L SS, 321 SS, 316L SS and Incoloy 825 are considered for the canister or overpack; 1020 carbon steel was selected for horizontal borehole liners, and a preliminary packing material selection is either compressed tuff or compressed tuff containing iron bearing smectite clay as a binder.
Date: September 1, 1983
Creator: Gregg, D.W. & O`Neal, W.C.
Partner: UNT Libraries Government Documents Department

Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

Description: Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.
Date: September 1, 1996
Creator: Ryskamp, J.M.; Adams, J.P.; Faw, E.M. & Anderson, P.A.
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

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

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

Diffusion of gases in solids: rare gas diffusion in solids; tritium diffusion in fission and fusion reactor metals. Final report

Description: Major results of tritium and rare gas diffusion research conducted under the contract are summarized. The materials studied were austenitic stainless steels, Zircaloy, and niobium. In all three of the metal systems investigated, tritium release rates were found to be inhibited by surface oxide films. The effective diffusion coefficients that control tritium release from surface films on Zircaloy and niobium were determined to be eight to ten orders of magnitude lower than the bulk diffusion coefficients. A rapid component of diffusion due to grain boundaries was identified in stainless steels. The grain boundary diffusion coefficient was determined to be about six orders of magnitude greater than the bulk diffusion coefficient for tritium in stainless steel. In Zircaloy clad fuel pins, the permeation rate of tritium through the cladding is rate-limited by the extremely slow diffusion rate in the surface films. Tritium diffusion rates through surface oxide films on niobium appear to be controlled by cracks in the surface films at temperatures up to 600/sup 0/C. Beyond 600/sup 0/C, the cracks appear to heal, thereby increasing the activation energy for diffusion through the oxide film. The steady-state diffusion of tritium in a fusion reactor blanket has been evaluated in order to calculate the equilibrium tritium transport rate, approximate time to equilibrium, and tritium inventory in various regions of the reactor blanket as a function of selected blanket parameters. Values for these quantities have been tabulated.
Date: September 1, 1976
Creator: Abraham, P. M.; Chandra, D.; Mintz, J. M.; Elleman, T. S. & Verghese, K.
Partner: UNT Libraries Government Documents Department

Reactor primary coolant system pipe rupture study progress report No. 36, 1 April--30 June 1976. [BWR]

Description: The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping, analytical and experimental efforts were started in 1965. The report summarizes the recent accomplshments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, and (b) studies directed at quantifying weld sensitization in Type-304 stainless steel.
Date: September 1, 1976
Partner: UNT Libraries Government Documents Department

Characterization of brazing alloys with stainless steels

Description: To simulate braze joints, qualitative x-ray mapping of the elemental interactions between brazing alloys and two common types of stainless steels was performed via the electron microprobe. In general both steels, Types 304L and 21- 6-9, react with a particular brazing alloy in a similar manner, the exceptions being the gold--copper brazing alloys which show deeper penetration into the 21-6- 9 stainless steel. (26 figures) (auth)
Date: September 19, 1975
Creator: Riefenberg, D. H.; Doyle, J. H.; Hillyer, R. F. & Bennett, W. S.
Partner: UNT Libraries Government Documents Department

Microscopic evaluation of low-temperature embrittlement in Type 308 stainless steel welds

Description: Effect of aging of type 308 stainless steel weld metal from 400 to 550C on microstructure was examined. Microstructural development was correlated with earlier results on mechanical properties that showed the ferrite-containing welds were prone to severe embrittlement when aged in this temperature range. The embrittlement was manifested by an increase in the ductile-brittle transition temperature and a drop in the upper-shelf energy. It was found that although the embrittlement over this aging temperature range was comparable, the microstructural changes that were responsible for the embrittlement were different at different temperatures. Embrittlement was caused by a combination of spinodal decomposition of ferrite, precipitation of M{sub 23}C{sub 6} carbide at the ferrite/austenite interface, and G-phase precipitation within the ferrite. Sigma phase formation at 550C may also be a contributing factor to the embrittlement.
Date: September 1, 1993
Creator: Vitek, J. M.; David, S. A. & Alexander, D. J.
Partner: UNT Libraries Government Documents Department

Hg removal from SRTC laboratory waste using an in-tank ion exchange probe

Description: An in-tank ion exchange column, described here, has been constructed to operate in the waste tanks at the Savannah River Technology Center (SRTC). The purpose of the column is to remove dissolved mercury from laboratory wastes and capture them on Duolite{trademark} GT-73 resin. This paper summarizes the results of operation of such a column on two batches of waste in one high activity tank and on one batch of waste in a low activity tank for the purpose of removing mercury from the liquid wastes. Sufficient mercury decontamination was demonstrated with the in-tank resin removal system, after twenty four hours of operation in each tank, to render the waste nonhazardous and enable shipment to the F-Area evaporator.
Date: September 25, 1992
Creator: Bibler, J. P. & DeGange, J. J.
Partner: UNT Libraries Government Documents Department

Heat exchanger, head and shell acceptance criteria

Description: Instability of postulated flaws in the head component of the heat exchanger could not produce a large break, equivalent to a DEGB in the PWS piping, due to the configuration of the head and restraint provided by the staybolts. Rather, leakage from throughwall flaws in the head would increase with flaw length with finite leakage areas that are bounded by a post-instability flaw configuration. Postulated flaws at instability in the shell of the heat exchanger or in the cooling water nozzles could produce a large break in the Cooling Water System (CWS) pressure boundary. An initial analysis of flaw stability for postulated flaws in the heat exchanger head was performed in January 1992. This present report updates that analysis and, additionally, provides acceptable flaw configurations to maintain defined structural or safety margins against flaw instability of the external pressure boundary components of the heat exchanger, namely the head, shell, and cooling water nozzles. Structural and flaw stability analyses of the heat exchanger tubes, the internal pressure boundary of the heat exchangers or interface boundary between the PWS and CWS, were previously completed in February 1992 as part of the heat exchanger restart evaluation and are not covered in this report.
Date: September 1, 1992
Creator: Lam, P. S. & Sindelar, R. L.
Partner: UNT Libraries Government Documents Department

The effect of yield strength on side-bonding upset welds

Description: During the course of 9{degree} tapered side-bonding resistance upset weld development at Mound, various studies have been conducted to evaluate the effect of yield strength on welds in 304L stainless steel. The results of these studies have concluded that at high yield strengths there may be a minor reduction in the length of Class 2 or better bond. Satisfactory welds have been produced with materials having yield strengths ranging from 36.0 to 141.0 ksi. However, when body yield strengths exceed 80.0 ksi a minor decrease in bond lengths begins. A significant inverse relationship between stem yield strength and bond length was shown to exist. 8 refs., 9 figs., 10 tabs.
Date: September 24, 1991
Creator: Miller, R.G. & Perkins, M.A.
Partner: UNT Libraries Government Documents Department

Hg removal from SRTC laboratory waste using an in-tank ion exchange probe

Description: An in-tank ion exchange column, described here, has been constructed to operate in the waste tanks at the Savannah River Technology Center (SRTC). The purpose of the column is to remove dissolved mercury from laboratory wastes and capture them on Duolite[trademark] GT-73 resin. This paper summarizes the results of operation of such a column on two batches of waste in one high activity tank and on one batch of waste in a low activity tank for the purpose of removing mercury from the liquid wastes. Sufficient mercury decontamination was demonstrated with the in-tank resin removal system, after twenty four hours of operation in each tank, to render the waste nonhazardous and enable shipment to the F-Area evaporator.
Date: September 25, 1992
Creator: Bibler, J.P. & DeGange, J.J.
Partner: UNT Libraries Government Documents Department