20 Matching Results

Search Results

Advanced search parameters have been applied.

Estimate of Legacy Tritium in Building 232-H Tritium Facility, Savannah River Site

Description: This report describes an estimate of how much tritium will be held up in those parts of the 232-H process that will remain in the building after deactivation The anticipated state of this tritium is also discussed. This information will be used to assess the radiological status of the deactivated facility.
Date: January 7, 2003
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Tritium Permeation Estimate from APT and CLWR-TEF Waste Packages

Description: The amount of tritium permeating out of waste containers has been estimated for the Accelerator Production of Tritium project (APT) and for the Commercial Light Water Reactor - Tritium Extraction Facility project (CLWR-TEF). The waste packages analyzed include the Aluminum, Window, Tungsten, Lead, and Steel packages for the APT project, and the overpack of extracted Tritium Producing Burnable Absorber Rods (TPBARs) for the CLWR-TEF project. All of the tritium contained in the waste was assumed to be available as a gas in the free volume inside the waste container at the beginning of disposal, and to then permeate the stainless steel waste container. From estimates of the tritium content of each waste form, the void or free volume of the package, disposal temperature and container geometry, the amount of tritium exiting the waste container by permeation was calculated. Two tritium permeation paths were considered separately: through the entire wall surface area and through the weld area only, the weld area having reduced thickness and significantly less surface area compared to the wall area. Permeation out of the five APT waste containers at 50 degrees Celsius is mainly through the welds, and at 100 degrees Celsius is through the permeation out of the entire wall surface area. The largest maximum offgas rate from an APT waste stream at 50 degrees Celsius (estimated disposal temperature) was 1.8E-6 Ci/year from the weld of the Window waste package, and the smallest maximum offgas rate was 3.7E-5 Ci/year from the weld of the Lead waste package. Permeation from the CLWR-TEF overpack at 40 degrees Celsius is mainly through the entire wall surface area, with a maximum offgas rate of 1.3E-5 Ci/year.
Date: March 18, 1999
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Decontaminating and Melt Recycling Tritium Contaminated Stainless Steel

Description: The Westinghouse Savannah River Company, Idaho National Engineering Laboratory, and several university and industrial partners are evaluating recycling radioactively contaminated stainless steel. The goal of this program is to recycle contaminated stainless steel scrap from US Department of Energy national defense facilities. There is a large quantity of stainless steel at the DOE Savannah River Site from retired heavy water moderated Nuclear material production reactors (for example heat exchangers and process water piping), that will be used in pilot studies of potential recycle processes. These parts are contaminated by fission products, activated species, and tritium generated by neutron irradiation of the primary reactor coolant, which is heavy (deuterated) water. This report reviews current understanding of tritium contamination of stainless steel and previous studies of decontaminating tritium exposed stainless steel. It also outlines stainless steel refining methods, and proposes recommendations based on this review.
Date: April 3, 1995
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Effects of 108 Days Tritium Exposure on UHMW-PE, PTFE, and Vespel(R)

Description: Samples of three polymers, Ultra-High Molecular Weight Polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE), also known as Teflon(R), and Vespel(R) polyimide were exposed to 1 atmosphere of tritium gas at ambient temperature for 108 days. Sample mass and size measurements to calculate density, spectra-colorimetry, dynamic mechanical analysis (DMA), and Fourier-transform infrared spectroscopy (FT-IR) were employed to characterize the effects of this exposure on these samples. This technical report is the first report from this research program.
Date: January 7, 2003
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Materials performance in prototype Thermal Cycling Absorption Process (TCAP) columns

Description: Two prototype Thermal Cycling Absorption Process (TCAP) columns have been metallurgically examined after retirement, to determine the causes of failure and to evaluate the performance of the column container materials in this application. Leaking of the fluid heating and cooling subsystems caused retirement of both TCAP columns, not leaking of the main hydrogen-containing column. The aluminum block design TCAP column (ABL block TCAP) used in the Advanced Hydride Laboratory, Building 773-A, failed in one nitrogen inlet tube that was crimped during fabrication, which lead to fatigue crack growth in the tube and subsequent leaking of nitrogen from this tube. The Third Generation stainless steel design TCAP column (Third generation TCAP), operated in 773-A room C-061, failed in a braze joint between the freon heating and cooling tubes (made of copper) and the main stainless steel column. In both cases, stresses from thermal cycling and local constraint likely caused the nucleation and growth of fatigue cracks. No materials compatibility problems between palladium coated kieselguhr (the material contained in the TCAP column) and either aluminum or stainless steel column materials were observed. The aluminum-stainless steel transition junction appeared to be unaffected by service in the AHL block TCAP. Also, no evidence of cracking was observed in the AHL block TCAP in a location expected to experience the highest thermal shock fatigue in this design. It is important to limit thermal stresses caused by constraint in hydride systems designed to work by temperature variation, such as hydride storage beds and TCAP columns.
Date: November 21, 1992
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Materials compatibility of hydride storage materials with austenitic stainless steels

Description: This task evaluated the materials compatibility of LaNi[sub 5-x]Al[sub x] (x= 0.3, 0.75) hydrides and palladium coated kieselguhr with austenitic stainless steel in hydrogen and tritium process environments. Based on observations of retired prototype hydride storage beds and materials exposure testing samples designed for this study, no materials compatibility problem was indicated. Scanning electron microscopy observations of features on stainless steel surfaces after exposure to hydrides are also commonly found on as-received materials before hydriding. These features are caused by either normal heat treating and acid cleaning of stainless steel or reflect the final machining operation.
Date: September 21, 1992
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Effect of nitrocarburizing on shape of titanium alloy parts

Description: Components are being developed for plutonium casting in support of Lawrence Livermore National Laboratory. A vendor used a proprietary process to grow a nitrocarburized surface layer on a titanium alloy shot sleeve to be used in a prototype die casting machine. The shot sleeve was significantly out-of-round upon return from the vendor and could not be used. Purpose of this study was to determine whether the shape change could have been caused by this surface treatment. Visual observation of disk and ring samples exposed first to surface treatment alone temperature and then the actual nitrocarburizing environment revealed no gross warping in either case. Dimension measurements of each sample before and after both the thermal treatment and the nitrocarburizing revealed no significant changes. Visual examination of the shot sleeve revealed a surface flaw likely made during handling after machining at SRS and before the part was nitrocarburized. The out-of-roundness of the shot sleeve could be related to the damage observed on the surface, but the possibility of warping during the nitrocarburizing cannot be excluded. Nitrocarburization should remain a candidate method to protect titanium alloys from molten metals.
Date: September 27, 1993
Creator: Clark, E. A.
Partner: UNT Libraries Government Documents Department

Hydroide Storage Vessel wall stress measurements

Description: Holographic Interferometry and strain gauge measurements were used to determine whether a prototype Hydride Storage Vessel (HSV) swelled while it was loaded in eleven stages with hydrogen. Bed swelling is inferred from deformation of the surface of the HSV. No swelling was detected, even after saturating the hydride material inside the HSV. The large chunky morphology of the titanium is likely responsible for the lack of wall stress. This morphology also implies that decay helium that remains in the titanium hydride (that is, helium that is not released as gas to the free volume) should not cause significant wall stresses when the HSV is used for long-term tritium storage. Holographic interferometry proved to be an extremely sensitive technique to measure swelling, having a detection limit of about 3 microns surface displacement.
Date: July 31, 1997
Creator: Clark, E.A. & Pechersky, M.J.
Partner: UNT Libraries Government Documents Department

Materials performance in prototype Thermal Cycling Absorption Process (TCAP) columns

Description: Two prototype Thermal Cycling Absorption Process (TCAP) columns have been metallurgically examined after retirement, to determine the causes of failure and to evaluate the performance of the column container materials in this application. Leaking of the fluid heating and cooling subsystems caused retirement of both TCAP columns, not leaking of the main hydrogen-containing column. The aluminum block design TCAP column (AHL block TCAP) used in the Advanced Hydride Laboratory, Building 773-A, failed in one nitrogen inlet tube that was crimped during fabrication, which lead to fatigue crack growth in the tube and subsequent leaking of nitrogen from this tube. The Third Generation stainless steel design TCAP column (Third generation TCAP), operated in 773-A room C-061, failed in a braze joint between the freon heating and cooling tubes (made of copper) and the main stainless steel column. In both cases, stresses from thermal cycling and local constraint likely caused the nucleation and growth of fatigue cracks. No materials compatibility problems between palladium coated kieselguhr (the material contained in the TCAP column) and either aluminum or stainless steel column materials were observed. The aluminum-stainless steel transition junction appeared to be unaffected by service in the AHL block TCAP. Also, no evidence of cracking was observed in the AHL block TCAP in a location expected to experience the highest thermal shock fatigue in this design. It is important to limit thermal stresses caused by constraint in hydride systems designed to work by temperature variation, such as hydride storage beds and TCAP columns.
Date: November 21, 1992
Creator: Clark, E. A.
Partner: UNT Libraries Government Documents Department

Type 304L stainless steel surface microstructure: Performance in hydride storage and acid cleaning

Description: The performance of stainless steel as the container in hydride storage bed systems has been evaluated, primarily using scanning electron microscopy. No adverse reaction between Type 304L stainless steel and either LaNi{sub 5{minus}x},Al{sub x}, or palladium supported on Kieselguhr granules (silica) during exposure in hydrogen was found in examination of retired prototype storage bed containers and special compatibility test samples. Intergranular surface ditching, observed on many of the stainless steel surfaces examined, was shown to result from air annealing and acid cleaning of stainless steel during normal fabrication. The ditched air annealed and acid cleaned stainless steel samples were more resistant to subsequent acid attack than vacuum annealed or polished samples without ditches.
Date: July 1, 1994
Creator: Clark, E. A.
Partner: UNT Libraries Government Documents Department

Materials compatibility of hydride storage materials with austenitic stainless steels

Description: This task evaluated the materials compatibility of LaNi{sub 5-x}Al{sub x} (x= 0.3, 0.75) hydrides and palladium coated kieselguhr with austenitic stainless steel in hydrogen and tritium process environments. Based on observations of retired prototype hydride storage beds and materials exposure testing samples designed for this study, no materials compatibility problem was indicated. Scanning electron microscopy observations of features on stainless steel surfaces after exposure to hydrides are also commonly found on as-received materials before hydriding. These features are caused by either normal heat treating and acid cleaning of stainless steel or reflect the final machining operation.
Date: September 21, 1992
Creator: Clark, E. A.
Partner: UNT Libraries Government Documents Department

Literature survey of tritiated waste characterization and disposal

Description: Characterizing, handling, and storing tritiated waste is challenging because of the physical and chemical properties of tritium. Tritium is soluble in many materials, including structural materials such as, stainless steel, structural steel, polymers, concrete and paints. Tritium permeates rapidly into these materials compared to other species, and so parts exposed to tritium are normally contaminated to some degree throughout the bulk. The relatively low kinetic energy of the {beta}-decay causes detecting tritium anywhere but very near the surface of materials to be impossible, because the {beta}-particle is absorbed by the material. Tritium readily exchanges with hydrogen in water vapor, and the resulting tritiated water can permeate polymers, concrete, oil, and the oxide surface films normally present on metals. Most of the tritium contamination in structural metals resides in the surface oxide film and in organic films at the surface, when metals are exposed to tritium at ambient temperature and pressure, whether the exposure is to gas or tritiated water. The most reliable method of assaying tritium is to dissolve samples in a proper liquid scintillant and use {beta}-scintillation counting. Other methods that require less time or are non-destructive (such as smear/counting) are significantly less reliable, but they can be used for routine waste characterization if sample dissolution/liquid scintillation counting is regularly employed to benchmark them.
Date: September 6, 1996
Creator: Clark, E. A.
Partner: UNT Libraries Government Documents Department

Feasibility study of acoustic emission monitoring of pinch welding tritium reservoir fill stems at the Savannah River Site

Description: A study was conducted to determine whether acoustic emission monitoring would be feasible in monitoring the solid-state resistance pinch weld used to seal tritium reservoirs at the Savannah River Site. Experiments were performed using a commercially available acoustic emission detection system, with a transducer mounted on a flat milled onto one of the pinch weld electrodes. Welds were made using a wide range of weld power, from very cold, with no metallurgical bond, to hot, with local fusion and excessive material injection into the tube bore. The tubes were drawn type 316L stainless steel. The welds were confined (anvils prevented material flow outward from the sides of the tube not being forced inward by the electrodes) and all were made using the same electrode force. The total number of ringdown counts for each weld was more correlated with weld power and bond length than total energy counts or total number of hits. The onset of large acoustic emission at higher weld power coincides with the injection of material into the tube bore, termed extrusion if arising from a solid state weld or spitting if arising from a weld with local fusion. Since large extrusions and spits, identified by radiography, cause rejection of production welds, a useful function of acoustic emission monitoring of pinch welding might be to detect the onset of extrusion or spitting. The low level of acoustic emission at production weld power levels (and below), the variability of acoustic emission at power levels causing extrusion and spitting, and the inability of acoustic emission to distinguish welds made with oxidized stems indicates that acoustic emission monitoring would not be a useful nondestructive evaluation of reservoir pinch welding at the Savannah River Site. 3 refs., 3 figs.
Date: January 1, 1990
Creator: Clark, E.A.
Partner: UNT Libraries Government Documents Department

Materials compatibility and wall stresses in hydride storage beds

Description: Hydrogen isotope handling and storage will be accomplished using solid-state hydride compounds at the Savannah River Site in the new Replacement Tritium Facility (RTF). The hydride powder is contained in a horizontal cylindrical vessel, and the combination of hydride powder, vessel, and associated heating and cooling facilities are termed in a hydride storage bed. The materials compatibility of the storage powder with the stainless steel vessel has been examined, and the stresses developed in the vessel due to expansion of the powder by absorbing hydrogen have been measured.
Date: January 1, 1991
Creator: Clark, E.A.; Dunn, K.A.; McKillip, S.T. & Bannister, C.E.
Partner: UNT Libraries Government Documents Department

Stress analysis of hydride bed vessels used for tritium storage

Description: A prototype hydride storage bed, using LaNi{sub 4.25}Al{sub 0.75} as the storage material, was fitted with strain gages to measure strains occurring in the stainless steel bed vessel caused by expansion of the storage powder upon uptake of hydrogen. The strain remained low in the bed as hydrogen was added, up to a bed loading of about 0.5 hydrogen to metal atom ratio (H/M). The strain then increased with increasing hydrogen loading ({approximately} 0.8 H/M). Different locations exhibited greatly different levels of maximum strain. In no case was the design stress of the vessel exceeded.
Date: January 1, 1991
Creator: McKillip, S.T.; Bannister, C.E. & Clark, E.A.
Partner: UNT Libraries Government Documents Department

Materials compatibility and wall stresses in hydride storage beds

Description: Hydrogen isotope handling and storage will be accomplished using solid-state hydride compounds at the Savannah River Site in the new Replacement Tritium Facility (RTF). The hydride powder is contained in a horizontal cylindrical vessel, and the combination of hydride powder, vessel, and associated heating and cooling facilities are termed in a hydride storage bed. The materials compatibility of the storage powder with the stainless steel vessel has been examined, and the stresses developed in the vessel due to expansion of the powder by absorbing hydrogen have been measured.
Date: December 31, 1991
Creator: Clark, E. A.; Dunn, K. A.; McKillip, S. T. & Bannister, C. E.
Partner: UNT Libraries Government Documents Department

Stress analysis of hydride bed vessels used for tritium storage

Description: A prototype hydride storage bed, using LaNi{sub 4.25}Al{sub 0.75} as the storage material, was fitted with strain gages to measure strains occurring in the stainless steel bed vessel caused by expansion of the storage powder upon uptake of hydrogen. The strain remained low in the bed as hydrogen was added, up to a bed loading of about 0.5 hydrogen to metal atom ratio (H/M). The strain then increased with increasing hydrogen loading ({approximately} 0.8 H/M). Different locations exhibited greatly different levels of maximum strain. In no case was the design stress of the vessel exceeded.
Date: December 31, 1991
Creator: McKillip, S. T.; Bannister, C. E. & Clark, E. A.
Partner: UNT Libraries Government Documents Department

SRS history and experience with palladium diffusers. Revision 1

Description: The Savannah River Site (SRS) has processed tritium in support of national defense programs since 1955. Palladium diffusers have been used extensively for separating hydrogen isotopes from inert gases (such as argon, helium, and nitrogen). In almost forty years of service, the design of the diffuser has been steadily improving. Several diffuser designs from different manufacturers have been evaluated at SRS. The operating experience gained from these designs together with failure analyses performed on failed units have led to several recommendations for improved diffuser designs and operating methods. This experience gained at SRS and the following recommendations form the basis of this report. Even though palladium diffuser technology has proven to be reliable, SRS has examined several alternative technologies over the past several years. This report will also review some of these promising alternatives.
Date: August 11, 1995
Creator: Clark, E.A.; Dauchess, D.A.; Heung, L.K.; Rabun, R.L. & Motyka, T.
Partner: UNT Libraries Government Documents Department

Experience with Palladium Diffusers in Tritium Processing

Description: Hydrogen isotopes are separated from other gases by permeation through palladium and palladium-silver alloy diffusers in the Tritium Facilities at the US Department of Energy Savannah River Site (SRS). Diffusers have provided effective service for almost forty years. This paper is an overview of the operational experience with the various diffuser types that have been employed at SRS. Alternative technologies being developed at SRS for purifying hydrogen isotopes are also discussed.
Date: January 27, 1995
Creator: Motyka, T.; Clark, E. A.; Dauchess, D. A.; Heung, L. K. & Rabum, R. L.
Partner: UNT Libraries Government Documents Department

Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

Description: The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.
Date: November 30, 1997
Creator: Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N. et al.
Partner: UNT Libraries Government Documents Department