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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 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

The effect of product form upon fatigue-crack growth behavior in Alloy 718: Additional results

Description: A previous study had characterized the fatigue-crack growth behavior of four wrought product forms (sheet, plate, bar and forging) from a single heat of Alloy 718 and concluded that there were no consistent trends in the crack growth rate results that could be attributed to product form variability. The present study adds one additional product form (gas-tungsten-arc weldments) from the same heat, and compares the behavior to that exhibited by the wrought product forms. Two different precipitation heat-treatments were employed at each of five test temperatures. 11 refs., 5 figs., 3 tabs.
Date: August 1, 1980
Creator: James, L.A.
Partner: UNT Libraries Government Documents Department

Fuel Storage Facility Final Safety Analysis Report. Revision 1

Description: The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.
Date: March 1, 1984
Creator: Linderoth, C.E.
Partner: UNT Libraries Government Documents Department

Sodium-to-gas leak-detection testing for FFTF: Letter 7754921

Description: The current status of the program is as follows: Work is in progress on Evaluation of Sodium Aerosol Generation Versus Leak Rate. Testing on Demonstration of Contact Leak Detector Performance in FFTF Reactor Inlet Guard Pipe has been completed and the final report is in preparation. Testing of Sodium Ionization Detector Response to Extraneous Materials is delayed until a current design Sodium Ionization Detector is available.
Date: November 11, 1977
Partner: UNT Libraries Government Documents Department

FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

Description: Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.
Date: May 1, 1981
Creator: Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A. & Daughtry, J.W.
Partner: UNT Libraries Government Documents Department

Selection of process parameters for sodium removal via the water vapor nitrogen process

Description: For the vapor phase of the WVN process the 160 to 190{sup 0}F temperature limit is shown to be well within the 145 to 208{sup 0}F minimum/maximum range. Decreasing process time by expansion of the temperature range is not expected to aid processing. The most productive area for improvement would be an increase in the water vapor concentration above the present 5% level. However, this would require confirmatory testing before approved use. The rinsing process was shown to be mainly controlled by component crevice geometry. Improvements in rinse time may be made by increasing the water temperature, but the concern over the caustic stress corrosion cracking will tend to limit the available increase. Although directed jets or sprays of rinse flows was not recommended, methods were suggested for conserving rinse water. Drying (as well as heating and cooling) of components was again shown to be constrained mostly by individual geometry and not processing parameters. A gas only, vacuum only, or a combination of the two modes were shown to be generally accepted methods. The hot gas only mode was recommended for its simplicity.
Date: July 1, 1976
Creator: Crippen, M.
Partner: UNT Libraries Government Documents Department

SIEX: a correlated code for the prediction of Liquid Metal Fast Breeder Reactor (LMFBR) fuel thermal performance

Description: The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperatures. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ``consistant`` set of material properties and correlated models provided by the code. The development of physically based correlations to model certain of the phenomana has resulted in a computer program which provides reliable estimates of thermal performance characteristics, yet requires a small amount of core storage and computer running time.
Date: January 1, 1974
Creator: Dutt, D.S. & Baker, R.B.
Partner: UNT Libraries Government Documents Department

Large-scale sodium-basalt concrete reaction test LSC-1

Description: The energy and hydrogen released from sodium-concrete reactions must be considered the analysis of beyond-design basis accidents for breeder reactors. Consequently, a large-scale sodium-basalt concrete reaction test was completed in the Large Sodium Fire Facility (LSFF) at the Hanford Engineering Development Laboratory (HEDL). 454 kg of sodium at 593{sup 0}C was spilled onto 0.84 m{sup 2} of basalt concrete 0.61 m deep containing two layers of reinforcing steel bar. From the data obtained, it was possible to complete a mass and energy balance for this test. The hydrogen generation and generation rate as functions of time for the duration of the test were determined. The major contribution to the chemical energy was energy associated with the formation of hydrogen (sodium-water reactions).
Date: June 1, 1981
Creator: McCormick, M.W.; Muhlestein, L.D.; Colburn, R.P. & Winkel, B.V.
Partner: UNT Libraries Government Documents Department

Post test valve examination of Sargent Globe valve

Description: The examination was made to observe and record any abnormalities caused by verification testing in liquid sodium which simulated 20 years of reactor service in FFTF and LMFBRs. The test program consisted of gas leakage measurements before and after a series of closing and opening actuations and thermal transient cycle exposures. External loads were used to simulate flow-induced vibration. Recommendations are given for applying these valves to future sodium systems. (23 figures) (DCL)
Date: March 1, 1977
Creator: Funk, C.W.
Partner: UNT Libraries Government Documents Department

PFR/Treat Safety Experiments: HEDL Transient Test Program Engineering Test Plan

Description: The purpose of the PFR/TREAT Safety Test Program is to obtain experimental data of fuel pin behavior during hypothetical, unprotected accidents for cores of large liquid metal cooled fast breeder reactors. The steady state and transient experiments, which will be performed under the joint program, are to be as prototypic of fast reactor performance as is possible. The specific objectives of this document are: (1) dictate the activities and responsibilities for the HEDL Transient Test Program; (2) specify the technical requirements for the CO4, CO5, CO6 and CO7 test train (SPTTs); and (3) specify the technical requirement for the CO6 and CO7 Single Pin Test Loops (SPTLs). Specific requirements for single pin loop experiments beyond CO7 and multi pin experiments will be covered in the addenda to this test plan.
Date: March 1, 1981
Creator: Hoffman, M.A.; Metcalf, I.L. & Myron, D.L.
Partner: UNT Libraries Government Documents Department

Alloy Development Program. Quarterly technical progress letter, October, November, December 1976

Description: Progress is reported in six chapters: swelling and creep, analytical studies (of irradiation effects), coolant compatibility (sodium), and status of EBR-II irradiation tests (one table). Materials studied include HT-9, 330 ss, Inconel 706, A-286, Nimonic PE16, Inconel 718, 310 ss, various developmental alloys, and 316 ss. Nickel ions as well as reactor irradiations were used in the studies of radiation effects. (212 figs., 54 tables). (DLC)
Date: April 1, 1977
Creator: Laidler, J.J. (comp.)
Partner: UNT Libraries Government Documents Department

Effect of heat treatment and heat-to-heat variations in the fatigue-crack growth response of Alloy 718. Part 2. Microscopic observation

Description: The microstructural aspects that influenced the room temperature and elevated temperature fatigue-crack propagation response of annealed, conventional, and modified heat-treated Alloy 718 were studied. Electron fractographic examination of Alloy 718 fatigue fracture surfaces revealed that operative crack growth mechanisms were dependent on heat treatment, heat-to-heat variations, temperature, and prevailing crack tip stress intensity level. In the low temperature regime (below 538{sup 0}C), all fracture surfaces exhibited a faceted appearance at low {Delta} levels, which is indicative of crystallographic fracture along intense inhomogeneous slip bands. The facets in the modified Alloy 718, however, were found to be rather poorly defined since the modified heat treatment tends to promote more homogeneous slip processes. Under progressively higher stress intensity levels, the room temperature and elevated temperature fatigue fracture surfaces exhibited striations, followed by a combination of striations and dimple rupture at the highest {Delta} values. Striation spacing measurements in all three heat-treated conditions were generally found to be in agreement with macroscopic growth rates at 24 and 538{sup 0}C. Under high temperature conditions (above 538{sup 0}C), evidence of intergranular fracture was also detected on the fatigue fracture surfaces, particularly at low stress intensity levels. This intergranular failure mechanism was found to be more extensive in the modified heat-treated Alloy 718. 17 figures.
Date: April 1, 1980
Creator: Mills, W.J. & James, L.A.
Partner: UNT Libraries Government Documents Department

FFTF transient overpower accident: a perspective

Description: This paper is a reflection on the current understanding of the unprotected transient overpower (TOP) accident, in order to place it in perspective with regard to FFTF core energetics. The experimental data base is addressed, wtih particular emphasis on the E and H-series data, and its relevance to axial failure location and hydraulic fuel sweepout is considered. It is shown that the only way in which TOP could lead to a sizeable energy release is if either total plugging takes place or a plug at the radial center of the subassemblies propagates to the hex can walls prior to neutronics shutdown. (DLC)
Date: February 1, 1975
Creator: Waltar, A.E.
Partner: UNT Libraries Government Documents Department

Sodium technology. Progress report, July-September 1980

Description: This report presents a quarterly summary of progress made in the areas of radioactivity control technology and sodium systems technology. Accomplishments during this period include: radionuclide trap operation in EBR-2; a 8000-h test of radionuclide deposition into 304 and 316 ss; radioactivity surveillance in FFTF HTS; inspection of deposition sampler from EBR-2; sodium frost tests; cold trap testing; effects of mesh packing on natural convection in cold trap crystallizer; and fuel failure monitoring in FFTF and EBR-2. (DLC)
Date: December 1, 1980
Creator: Atwood, J.M. (comp.)
Partner: UNT Libraries Government Documents Department

Alternative Fuel Cycle Evaluation Program. Volume IV. International Fuel Service Center evaluation. Revision 1

Description: This Alternative Fuel Cycle Evaluation Program (AFCEP) study presents the technical, economic and social aspects of the International Fuel Service Center (IFSC) as an institutional approach to nuclear fuel cycle development and is provided in support of the Nonproliferation Alternative Systems Assessment program (NASAP). Four types of IFSCs are described and evaluated in terms of three different twenty-year nuclear growth scenarios. Capital costs for each IFSC and comparable dispersed facility costs are discussed. Finally, the possible impact of each scenario and IFSC on the environmental and socio-economic structure is examined. 14 refs., 33 figs., 15 tabs.
Date: November 1, 1979
Creator: Jacobson, L D
Partner: UNT Libraries Government Documents Department

Fast breeder reactor fuel pins: Revision 1984

Description: This standard establishes the requirements for fuel pins to be used in FBR fuel assemblies. Fuel pins consist of mixed uranium-plutonium oxide fuel pellets clad with Type 316 stainless steel or other purchaser specified alloy steel.
Date: January 1, 1984
Partner: UNT Libraries Government Documents Department