Mechanical response of FFTF reference and P1 cladding tubes under transient heating [electronic resource].
- Published:
- Argonne, Ill. : Argonne National Laboratory, 1977.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy. - Physical Description:
- Pages: 13 : digital, PDF file
- Additional Creators:
- Argonne National Laboratory and United States. Department of Energy. Office of Scientific and Technical Information
Access Online
- Restrictions on Access:
- Free-to-read Unrestricted online access
- Summary:
- Burst tests of Type 316 stainless steel cladding tube samples subjected to increasing temperature and relatively constant internal pressure were conducted to assist in the pretest analysis of the P1 experiment performed in the Sodium Loop Safety Facility. This paper reports and analyzes the burst test results and those of subsequent transient heating work. The use of a modified extensometer in obtaining mechanical response data for stainless steel in the high temperature range is illustrated, some of such data is provided, and the potential of further experiments and analysis is indicated. Tubing of the same design as Fast Flux Test Facility (FFTF) cladding (20% cold worked, 0.230 in. OD, 15 mil wall) was tested as-received and after annealing or electrolytic thinning. P1 tubing (38% cold worked, 0.230 in. OD, 10 mil wall) was tested before and after aging under conditions anticipated in the P1 reactor experiment. The P1 cladding was designed to simulate FFTF tubing that had experienced irradiation embrittlement and attack by cesium oxide and sodium impurities.
- Report Numbers:
- E 1.99:conf-770807-43
conf-770807-43 - Subject(s):
- Other Subject(s):
- Fftf Reactor
- Fuel Cans
- Ruptures
- Stainless Steel-316
- Thermal Stresses
- Very High Temperature
- Alloys
- Chromium Alloys
- Chromium Steels
- Chromium-Nickel Steels
- Corrosion Resistant Alloys
- Epithermal Reactors
- Failures
- Fast Reactors
- Heat Resisting Alloys
- Iron Alloys
- Iron Base Alloys
- Liquid Metal Cooled Reactors
- Molybdenum Alloys
- Nickel Alloys
- Reactors
- Research And Test Reactors
- Research Reactors
- Sodium Cooled Reactors
- Stainless Steels
- Steels
- Stresses
- Test Reactors
- Note:
- Published through SciTech Connect.
01/01/1977.
"conf-770807-43"
4. international conference on structural mechanics in reactor technology, San Francisco, California, USA, 15 Aug 1977.
Ariman, T.; Lepacek, B.E.; Youngahl, C.A. - Funding Information:
- W-31-109-ENG-38
View MARC record | catkey: 14803895