Evaluation of sustained load effects in reactor pressure vessels by means of intermediate-scale flawed vessel tests. [BWR; PWR]

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Application of results of fracture tests of flawed vessels to full-scale reactor pressure vessels must take into account differences in scale, which may affect both the stress field around the flaw and the effective material toughness, as well as differences in the loading system which affect crack behavior through load variations. The intermediate test vessels (ITV's) of the Heavy Section Steel Technology program were designed with essentially full-scale thickness so that fracture initiation behavior could be demonstrated with minimal uncertainty associated with scale. In most of the ITV tests the stiffness of the hydraulic loading system relative to that of ... continued below

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Pages: 27

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Bryan, R. H. January 1, 1976.

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Application of results of fracture tests of flawed vessels to full-scale reactor pressure vessels must take into account differences in scale, which may affect both the stress field around the flaw and the effective material toughness, as well as differences in the loading system which affect crack behavior through load variations. The intermediate test vessels (ITV's) of the Heavy Section Steel Technology program were designed with essentially full-scale thickness so that fracture initiation behavior could be demonstrated with minimal uncertainty associated with scale. In most of the ITV tests the stiffness of the hydraulic loading system relative to that of a real reactor system was of no consequence, since the terminal fracture was rapid and extensive. Two of the ITV tests resulted in leak without burst, however, which suggested the study of the behavior of intermediate test vessels under sustained load, as would be obtained in a large pressurized-water or boiling-water reactor system. Consequently the test of ITV-7 has been repeated with a flaw and test conditions nearly identical to the original flaw and conditions but with the load imposed pneumatically and the flawed region covered with a patch to retard or prevent leakage at the time of rupture. The results of the tests suggest that the demonstrations of leak without burst in the intermediate vessel tests, both hydraulic and pneumatic, are applicable to the evaluation of the behavior of reactor pressure vessels with similar flaw geometries under sustained load. The two vessels that ruptured in this way withstood pressures 2.15 to 2.74 times design pressure. These test pressures are above the ASME Boiler and Pressure Vessel Code allowable pressures for faulted conditions.

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Pages: 27

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  • 4. water reactor safety research information meeting, Washington, DC, USA, 27 Sep 1976

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  • Report No.: CONF-760954-5
  • Grant Number: W-7405-ENG-26
  • Office of Scientific & Technical Information Report Number: 7324716
  • Archival Resource Key: ark:/67531/metadc1444973

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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  • January 1, 1976

Added to The UNT Digital Library

  • Feb. 10, 2019, 8:45 p.m.

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  • Feb. 20, 2019, 5:09 p.m.

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Bryan, R. H. Evaluation of sustained load effects in reactor pressure vessels by means of intermediate-scale flawed vessel tests. [BWR; PWR], article, January 1, 1976; Tennessee. (https://digital.library.unt.edu/ark:/67531/metadc1444973/: accessed March 26, 2019), University of North Texas Libraries, Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.