skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Evaluation of material wear and self-welding in sodium-cooled reactor systems

Abstract

Laboratory and prototype component tests were made to determine the degree of adhesion and self-welding of various bearing-couple materials in liquid sodium. Tests on these materials were also run in nitrogen and in argon. Temperatures reached were from 500 to 1000 deg F, and the static loads applied on the bearing surfaces were up to 10,000 psi. Results of the tests demonstrated that the harder materials generally showed better wear resistance than did the softer materials. However, each material has a service life depending upon such factors as: temperature, unit loads, nature of service, and environment. (auth)

Authors:
;
Publication Date:
Research Org.:
Atomic Power Development Associates, Inc., Detroit, MI (United States)
Sponsoring Org.:
US Atomic Energy Commission (AEC)
OSTI Identifier:
4282062
Report Number(s):
APDA-126
NSA Number:
NSA-13-004689
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
METALLURGY AND CERAMICS; ADHESION; ARGON; BEARINGS; ENVIRONMENT; HARDNESS; INERT GASES; LABORATORY EQUIPMENT; LIQUID METAL COOLANT; MATERIALS TESTING; NITROGEN; PRESSURE; REACTORS; SODIUM; STABILITY; SURFACES; TEMPERATURE; WEAR; WELDING

Citation Formats

Leeser, D. O., and Williams, R. C. Evaluation of material wear and self-welding in sodium-cooled reactor systems. United States: N. p., 1958. Web. doi:10.2172/4282062.
Copy to clipboard
Leeser, D. O., & Williams, R. C. Evaluation of material wear and self-welding in sodium-cooled reactor systems. United States. doi:10.2172/4282062.
Copy to clipboard
Leeser, D. O., and Williams, R. C. Fri . "Evaluation of material wear and self-welding in sodium-cooled reactor systems". United States. doi:10.2172/4282062. https://www.osti.gov/servlets/purl/4282062.
Copy to clipboard
@article{osti_4282062,
title = {Evaluation of material wear and self-welding in sodium-cooled reactor systems},
author = {Leeser, D. O. and Williams, R. C.},
abstractNote = {Laboratory and prototype component tests were made to determine the degree of adhesion and self-welding of various bearing-couple materials in liquid sodium. Tests on these materials were also run in nitrogen and in argon. Temperatures reached were from 500 to 1000 deg F, and the static loads applied on the bearing surfaces were up to 10,000 psi. Results of the tests demonstrated that the harder materials generally showed better wear resistance than did the softer materials. However, each material has a service life depending upon such factors as: temperature, unit loads, nature of service, and environment. (auth)},
doi = {10.2172/4282062},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Aug 01 00:00:00 EDT 1958},
month = {Fri Aug 01 00:00:00 EDT 1958}
}
Copy to clipboard
Technical Report:
View Technical Report (21.90 MB)
DOI:  10.2172/4282062
Save / Share:
Export Metadata
Save to My Library
You must Sign In or Create an Account in order to save documents to your library.

  • ; ; - Journal of Basic Engineering (U.S.) Divided into J. Eng. Mater. Technol. and J. Fluids Eng.
    Development Associates Inc., Detroit). J. Basic Eng. Laboratory and prototype component tests were made to determine the degree of adhesion and self- welding of various bearing-couple materials in liquid Na. A few tests were run in N/sub 2/ and Ar. Temperatures were from 500 to 1000 deg F and the stutic loads on the bearing surfaces were up to 10,000 psi. The harder materials showed better wear resistance than did softer materials. (auth)

  • Material inspection and welding specifications are presented for various parts of both the intermediate heat exchanger and steam generator. Tables are included that indicate the applicable parts and assemblies to which these specifications apply. For other parts, where the material requirements are not severe, the ASTM or other indicated specifications are applicable. (J.R.D.)

  • A series of prototypic steam generator 2-{1/4} Cr-1 Mo tube/alloy 718 tube support plate wear tests were conducted in direct support of the Westinghouse Nuclear Components Division -- Breeder Reactor Components Project Large Scale steam Generator design. The initial objective was to verify the acceptable wear behavior of softer, over-aged'' alloy 718 support plate material. For all interfaces under all test conditions, resultant wear damage was adhesive in nature with varying amounts of 2-{1/4} Cr-1 Mo tube material being adhesively transferred to the alloy 718 tube supports. Maximum tube wear depths exceeded the initially established design allowable limit of 127more » {mu}m (.005 in.) at 17 of the 18 interfaces tested. A decrease in contact stresses produced acceptable tube wear depths below a readjusted maximum design allowable value of 381 {mu}m (.015 in.). Additional conservatisms associated with the simulation of a 40-year lifetime of rubbing in a one-week laboratory test provided further confidence that the 381 {mu}m maximum tube wear allowance would not be exceeded in service. Softer, over-aged'' alloy 718 material was found to produce slightly less wear damage on 2-{1/4} Cr-1 Mo tubing than fully age hardened material. Also, air formed oxide films on the alloy 718 reduced initial tube wear and delayed the onset of adhesive surface damage. However, at high surface stress levels, these films were not sufficiently stable to provide adequate long term protection from adhesive wear. The results of the present work and those of previous test programs suggest that the successful in-sodium tribological performance of 2-{1/4} Cr-1 Mo/alloy 718 rubbing couples is dependent upon the presence of lubricative surface films, such as oxides and/or surface reaction or deposition products. 11 refs., 13 figs., 4 tabs.« less

  • With the purpose of investigating the self-welding tendency of structural materials in the light of the sodium cooled fast breeder reactor to be built, static load tests in sodium were carried out on these materials. Besides austenitic materials also a stabilized ferritic steel and a nickel base alloy as well as several hard facing and hard metal alloys were used in the tests. The test temperatures were 580 and 700 deg C, the compressive load ranged up to 5 kp/mm2, and the tests lasted for 168 hours. The oxygen content of the cold-trapped sodium was nearly 20 ppm. The resultsmore » show a more or less pronounced tendency of all the materials to weld together. While the austenitic and ferritic steels showed a marked tendency to weld together at 700 deg C, the behavior of the hard facing and the hard metal alloys is not uniform. The nickel base alloy, which up to now has been investigated in one experiment only, does not weld together - with one exception. Micrographs made of some different material combinations seem to indicate that the welding was caused by recrystallization after previous strong deformation of the contact region (auth)« less

  • ; ; ; ...
    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na 2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO 2 (sCO 2) Brayton cycle power conversion system. In SFRs,more » Na 2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na 2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na 2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na 2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO 2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test section and provided with Na 2O deposition prediction capabilities, was developed. This state-of-the-art computational tool incorporates a first-principles Na 2O deposition model developed by ANL, and combines it with predictive capabilities for the spatial and temporal variation of temperature, velocity, dissolved oxygen concentration, and wall temperature under flowing sodium conditions. The CFD model was validated under no-deposition conditions using experimental data collected with the SPPL, demonstrating the model’s capability to predict the thermal-hydraulics of the SPPL test section within the measurement uncertainty characterizing the SPPL instrumentation. The model’s deposition prediction capability was not, however, validated as the SPPL could not be operated under plugging conditions during the project, resulting in the lack of deposition data with adequate pedigree for a CFD model validation. Two novel diagnostic techniques to detect and characterize Na 2O deposits, i.e. Ultrasonic Time Domain Reflectometry (UTDR) and Potential Drop (PD) techniques, were developed to ultimately assist in the validation effort under plugging conditions, which can be performed once the SPPL becomes operational. This development effort consisted first in demonstrating, analytically and/or computationally, the capability of these techniques to diagnose Na 2O deposits inside of small channels (particularly the deposit’s thickness), and subsequently in the fabrication and testing of prototypical UTDR and PD instrumentation. The testing, performed on mockups of the SPPL test section, demonstrated the capability of these techniques to detect and characterize material discontinuities like those induced by sodium oxide deposition on stainless steel channel walls. Because of the mentioned impossibility to run the SPPL in a plugging mode, the developed instrumentation could not be tested in-situ, i.e. at the SPPL while deposits are being formed inside of the SPPL test section. Recommended future work includes a possible enhancement in the CFD modeling technique and installation of the developed UTDR and PD instrumentation on the test section, followed by plugging tests to be conducted with the SPPL. The installation of the UTDR and PD diagnostic instrumentation on the SPPL test section will allow collection of Na 2O deposition data after the onset of deposition to nearly complete channel plugging, which can ultimately be used for the validation of the CFD model.« less