Boiling-water reactor internals aging degradation study
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Published by Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Supt. of Docs., U.S. G.P.O. [distributor] in Washington, DC .
Written in English


  • Boiling water reactors -- Materials -- Deterioration.,
  • Nuclear pressure vessels -- Materials -- Deterioration.

Book details:

Edition Notes

Other titlesBoiling water reactor internals aging degradation study.
Statementprepared by K.H. Luk.
ContributionsU.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering., Oak Ridge National Laboratory.
The Physical Object
Paginationxiv, 43 p.
Number of Pages43
ID Numbers
Open LibraryOL14692584M

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  Description In a comprehensive and lucid manner this book presents an understanding of the aging degradation of the major pressurized and boiling water reactor structures and components. The design and fabrication of each structure or component is briefly described followed by information on the associated Edition: 1. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC).Cited by: 1. A report by NRC published in confirmed that age-related degradation in BWRs will damage or destroy many vital safety-related components inside the reactor vessel before the forty year license expires. The NRC report states "Failure of internals could create conditions that may challenge the integrity the reactor primary containment systems.".

In order to provide a scientific basis for proposed life extension of current light water reactors, the radiation-induced degradation of stainless steel reactor internals will be discussed. A brief review of the basic radiation damage effects in stainless steels at LWR relevant conditions will be by: in use: pwr reactor internals aging management. ISSUE STATEMENT. Reactor pressure vessel internal components in pressurized water reactors may be affected by age-related degradation. effects. These include general material effects such as wear, fatigue and stress corrosion cracking, as well as irradiation-. • Synergistic Effects in Degradation • Synergetic effects of thermal aging and irradiation aging on degradation of stainless steel welds in reactor internals • MDM, Materials Information Portal and Materials Handbook • MDM Revision-3 has been published in May (EPRI Report - . •H 2O as coolant and moderator • Pressure in water/steam cycle: 70 bar (7 MPa) • Boiling of water in the core • Temperature about ºC • Steam transferred directly from core to turbine generator after passing steam/water separator • Average power density in core: 50 kW/litre • Burn-up: ca. MWd/t U • Thermal net efficiency: % BWR Basics.

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) provides an integrated approach for managing materials-related degradation issues in reactor coolant system components in boiling water reactors. The program assesses all facets of operation, maintenance, and repair to develop reliable and cost-effectiveFile Size: 67KB. @article{osti_, title = {Pressurized-water reactor internals aging degradation study. Phase 1}, author = {Luk, K H}, abstractNote = {This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and Cited by: 3. Get this from a library! Boiling-water reactor internals aging degradation study: phase 1. [K H Luk; U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering.; Oak Ridge National Laboratory.]. “Beginning January 1, , the EPRI Materials Degradation and Aging Action Plan Committee has the principal role for overseeing industry activities related to primary system materials and the continuing commitment to the Industry Materials Initiative (NEI ).