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The TRIO Experiment

Description: The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.
Date: September 1984
Creator: Clemmer, Robert G.
Partner: UNT Libraries Government Documents Department

Experimental basis for parameters contributing to energy dissipation in piping systems

Description: The paper reviews several pipe testing programs to suggest the phenomena causing energy dissipation in piping systems. Such phenomena include material damping, plasticity, collision in gaps and between pipes, water dynamics, insulation straining, coupling slippage, restraints (snubbers, struts, etc.), and pipe/structure interaction. These observations are supported by a large experimental data base. Data are available from in-situ and laboratory tests (pipe diameters up to about 20 inches, response levels from milli-g's to responses causing yielding, and from excitation wave forms including sinusoid, snapback, random, and seismic). A variety of pipe configurations have been tested, including simple, bare, straight sections and complex lines with bends, snubbers, struts, and insulation. Tests have been performed with and without water and at zero to operating pressure. Both light water reactor and LMFBR piping have been tested.
Date: January 1, 1985
Creator: Ibanez, P. & Ware, A.G.
Partner: UNT Libraries Government Documents Department

Application of microprocessor based controller in the Breeder Reactor Program

Description: This paper treats Argonne National Laboratory's experience using microprocessor based controllers presently in use on several control loops within the EBR-II reactor facility as well as tests being performed by these controllers. Also included is a discussion of the expandability, modularity, range of capabilities and higher level functions possible using such equipment.
Date: January 1, 1985
Creator: Messick, N.C. Lukas, M.P.
Partner: UNT Libraries Government Documents Department

Application of aerosol technology in LMFBR design

Description: Aerosol technology is applied in several areas in the safety assessment of liquid metal fast breeder reactors. This paper discusses the application of this technology in the assessment of the Clinch River Breeder Reactor Plant. The importance of considering aerosol effects is discussed for sodium fires, the assessment of site suitability and the assessment of the consequences of accidents beyond the design base such as hypothetical core disruptive accidents. Areas in which further development work could have the most impact are indicated.
Date: January 1, 1980
Creator: Strawbridge, L.E. & Hemmerle, E.H.
Partner: UNT Libraries Government Documents Department

Effect of heat treatment and heat-to-heat variations in the fatigue-crack growth response of Alloy 718

Description: The fatigue-crack growth behavior of seven heats of Alloy 718 was studied at five different test temperatures. These seven heats represented at least four different producers, four different product forms, two melt practices, and most of the heat were tested in two different heat-treated conditions. Heat-to-heat variations were noted; these were most obvious in material given the conventional heat-treatment. 8 figs., 5 tabs.
Date: April 1, 1980
Creator: James, L.A. & Mills, W.J.
Partner: UNT Libraries Government Documents Department

Effect of heat treatment upon the fatigue-crack growth behavior of Alloy 718 weldments

Description: The microstructural features that influenced the room and elevated temperature fatigue-crack growth behavior of as-welded, conventional heat-treated, and modified heat-treated Alloy 718 GTA weldments were studied. Electron fractographic examination of fatigue fracture surfaces revealed that operative fatigue mechanisms were dependent on microstructure, temperatures and stress intensity factor. All specimens exhibited three basic fracture surface appearances at temperatures up to 538{degrees}C: crystallographic faceting at low stress intensity range ({Delta}K) levels, striation, formation at intermediate values, and dimples coupled with striations in the highest ({Delta}K) regime. At 649{degrees}C, the heat-treated welds exhibited extensive intergranular cracking. Laves and {delta} particles in the conventional heat-treated material nucleated microvoids ahead of the advancing crack front and caused on overall acceleration in crack growth rates at intermediate and high {Delta}K levels. The modified heat treatment removed many of these particles from the weld zone, thereby improving its fatigue resistance. The dramatically improved fatigue properties exhibited by the as-welded material was attributed to compressive residual stresses introduced by the welding process. 19 refs., 16 figs.
Date: May 1, 1981
Creator: Mills, W.J. & James, L.A.
Partner: UNT Libraries Government Documents Department

Effect of nonuniform inlet air flow on air-cooled heat-exchanger performance

Description: Blowers used to propel air across tube bundles generate a non-uniform flow field due to their construction details. A formalism to evaluate heat transfer degradation due to non-uniform airflow has been developed. Certain symmetry relations for cross flowheat exchangers, heretofore unavailable in the open literature, have been derived. The solution presented here was developed to model a 4 tube pass air blast heat exchanger for the Clinch River Breeder Reactor Plant Project. This case is utilized to show how this method can be used as a design tool to select the most suitable blower construction for a particular application. A numerical example is used to illustrate the salient points of the solution.
Date: January 1, 1983
Creator: Soler, A.I.; Singh, K.P. & Ng, T.L.
Partner: UNT Libraries Government Documents Department

Effect of pipe insulation losses on a loss-of-heat sink accident for an LMR

Description: The efficacy of pipe radiation losses as a heat sink during LOHS in a loop-type LMR plant is investigated. The Super System Code (SSC), which was modified to include pipe radiation losses, was used to simulate such an LOHS in an LMR plant. In order to enhance these losses, the pipes were assumed to be insulated by rock wool, a material whose thermal conductivity increases with increasing temperature. A transient was simulated for a total of eight days, during which the coolant temperatures peaked well below saturation conditions and then declined steadily. The coolant flow rate in the loop remained positive throughout the transient.
Date: January 1, 1985
Creator: Horak, W.C.; Guppy, J.G. & Wood, P.M.
Partner: UNT Libraries Government Documents Department

End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

Description: Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.
Date: October 1, 1987
Creator: Richardson, K.D.
Partner: UNT Libraries Government Documents Department

Emissivity of sodium wetted and oxidized Type 304 stainless steel

Description: The emissivity of sodium wetted and oxidized Type 304 stainless steel was determined to provide data for calculating the heat flow through Liquid Metal Fast Breeder Reactor (LMFBR) reflector plates, located above the sodium pool, to the reactor closure head. An emissivity experiment using a Type 304 stainless steel specimen was performed in an inerted glovebox. Relatively high oxygen concentrations of 10,000 and 50 vppm were used in the argon/oxygen mixtures to reduce reaction time. Following wetting and oxidation, the specimen was heated to a maximum temperature of 450/sup 0/C and the emissivity of the oxidized coating was calculated. Results indicate that the emissivity of the coating ranged from 0.55 to 0.92.
Date: January 1, 1980
Creator: Haines, N.L.; Craig, R.E.; Forsyth, D.R. & Novendstern, E.H.
Partner: UNT Libraries Government Documents Department

Effect of heat treatment upon the fatigue-crack growth behavior of Alloy 718 weldments

Description: Gas-tungsten-arc weldments in Alloy 718 were studied in fatigue-crack growth test conducted at five temperatures over the range 24--649{degree}C. In general, crack growth rates increased with increasing temperature, and weldments given the conventional'' post-weld heat-treatment generally exhibited crack growth rates that were higher than for weldments given the modified'' (INEL) heat-treatment. Limited testing in the as-welded condition revealed crack growth rates significantly lower than observed for the heat-treated cases, and this was attributed to residual stresses. Three different heats of filler wire were utilized, and no heat-to-heat variations were noted. 23 refs., 9 figs., 6 tabs.
Date: May 1, 1981
Creator: James, L.A. & Mills, W.J.
Partner: UNT Libraries Government Documents Department

Evaluation of fuel release rate and mechanism tests under RBCB conditions. [LMFBR]

Description: This task includes theoretical evaluation of fuel/fission product release behavior from failed LMFBR fuel elements as well as an on-going experimental investigation of the mechanism of oxide fuel dispersal into flowing liquid sodium. The primary objectives of this work are to develop a fuel source term that can be used in predictive models for primary heat transfer system contamination and to understand the separate influences of important system variables (such as flow rate, oxygen impurity level) on this source term. The present report is written in two parts: the first, in condensed form, is an updated evaluation of fuel (U,Pu) and fission product release data, and the second describes the current status of supporting experimental work at General Electric's Vallecitos Laboratory.
Date: September 1, 1981
Creator: Adamson, M.G.
Partner: UNT Libraries Government Documents Department

Evaluation of clamp effects on LMFBR piping systems

Description: Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness.
Date: January 1, 1980
Creator: Jones, G.L.
Partner: UNT Libraries Government Documents Department

Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR. [LMFBR]

Description: In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone.
Date: January 1, 1983
Creator: Chawla, T.C.; Pedersen, D.R.; Leaf, G. & Minkowycz, W.J.
Partner: UNT Libraries Government Documents Department

Energy Economic Data Base (EEDB) Program. Phase III. Final report and third update

Description: The objective of the USDOE EEDB Program is to provide periodic updates of technical and cost (capital, fuel and operating and maintenance) information of significance to the US Department of Energy. This information is intended to be used by USDOE in evauating and monitoring US civilian nuclear power programs, and to provide them with a consistent means of evaluating the nuclear option and proposed alternatives. The data tables, which make up the bulk of the report, are updated to January 1, 1980. The data in these tables and in the backup data file supercede the information presented in the Second Update (1979). Where required, new descriptive information is added in the text to supplement the data tables.
Date: July 1, 1981
Partner: UNT Libraries Government Documents Department

Light-water breeder reactors: preliminary safety and environmental information document. Volume III

Description: Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter.
Date: January 1, 1980
Partner: UNT Libraries Government Documents Department

Modularization and nuclear power. Report by the Technology Transfer Modularization Task Team

Description: This report describes the results of the work performed by the Technology Transfer Task Team on Modularization. This work was performed as part of the Technology Transfer work being performed under Department of Energy Contract 54-7WM-335406, between December, 1984 and February, 1985. The purpose of this task team effort was to briefly survey the current use of modularization in the nuclear and non-nuclear industries and to assess and evaluate the techniques available for potential application to nuclear power. A key conclusion of the evaluation was that there was a need for a study to establish guidelines for the future development of Light Water Reactor, High Temperature Gas Reactor and Liquid Metal Reactor plants. The guidelines should identify how modularization can improve construction, maintenance, life extension and decommissioning.
Date: June 1, 1985
Partner: UNT Libraries Government Documents Department

Robotics and nuclear power. Report by the Technology Transfer Robotics Task Team

Description: A task team was formed at the request of the Department of Energy to evaluate and assess technology development needed for advanced robotics in the nuclear industry. The mission of these technologies is to provide the nuclear industry with the support for the application of advanced robotics to reduce nuclear power generating costs and enhance the safety of the personnel in the industry. The investigation included robotic and teleoperated systems. A robotic system is defined as a reprogrammable, multifunctional manipulator designed to move materials, parts, tools, or specialized devices through variable programmed motions for the performance of a variety of tasks. A teleoperated system includes an operator who remotely controls the system by direct viewing or through a vision system.
Date: June 1, 1985
Partner: UNT Libraries Government Documents Department

SCTI chemical leak detection test plan

Description: Tests will be conducted on the CRBRP prototype steam generator at SCTI to determine the effects of steam generator geometry on the response of the CRBRP chemical leak detection system to small water-to-sodium leaks in various regions of the steam generator. Specifically, small injections of hydrogen gas (simulating water leaks) will be made near the two tubesheets, and the effective transport times to the main stream exit and vent line hydrogen meters will be measured. The magnitude and time characteristics of the meters' response will also be measured. This information will be used by the Small Leak Protection Base Program (SG027) for improved predictions of meter response times and leak detection sensitivity.
Date: October 12, 1981
Partner: UNT Libraries Government Documents Department

Supplementary information on Series II Test A-5 test conditions

Description: Test planning and test preparation for LLTR Series II Test A-5 are continuing at GE and ETEC. The main objective for this test is to obtain data on the type and magnitude of steam tube blowout failures resulting from worst case leak conditions under the superheater hot standby condition (i.e., the plant condition considered most susceptible to steam tube blowout failures from wastage/overheating). A corollary objective is to add large quantities of steam (i.e. approx. 330 lbs) to simulate the amount that could be added in a plant system (such as CRBRP) before pressure would build up in the intermediate Heat Transport System to blow the expansion tank rupture disc. This report recommends the preferred method for operating the LLTR primary/secondary steam systems for Test A-5. i.e., Option 5 - Common Supply Tanks for Primary and Secondary Systems.
Date: November 16, 1981
Partner: UNT Libraries Government Documents Department

Semi-annual report, summary of Rockwell International - Energy Systems Group contribution. [LMFBR]

Description: In the decontamination task, emphasis during this report period was on determining the effect of various sensitization conditions on the metal removal rate for various stainless steel alloys, using the reference decontamination process. An accurate prediction of metal removal rate is necessary to determine the optimum process time. Results to date indicate that alloys exposed to sodium at reactor conditions show a higher metal removal rate than the same types of alloys exposed to equivalent time-temperature conditions in a vacuum. Other decontamination work included preparations for determining the effect of the reference decontamination process on hardfacing alloys and assistance in design modifications of the Clinch River Reactor decontamination system. In the program task to develop an evaporative process to remove sodium from a full-size fuel subassembly, progress was made in the final assembly of the evaporation system. The test article, a fuel subassembly, was modified by adding heaters to simulate gamma heating. Thermocouples were attached at suitable internal and external positions. The vacuum containment system was assembled and leak checked. The assembly of the system is progressing on schedule and will be conpleted during the next report period.
Date: January 1, 1981
Partner: UNT Libraries Government Documents Department