Published April 1997 | Version v1
Report Restricted

Studies of the steam generator degraded tubes behavior on BRUTUS test loop

Creators

  • 1. EDF/DER/MTC, Moret Sur Loing (France)
  • 2. EDF/EPN/DMAINT, Paris (France)

Description

Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture

Files

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Additional details

Publishing Information

Imprint Title
Proceedings of the seminar on leak before break in reactor piping and vessels
Imprint Pagination
773 p.
Journal Page Range
p. 741-748.
Report number
NUREG/CP--0155

Conference

Title
Seminar on leak before break in reactor piping and vessels.
Dates
9-11 Oct 1995.
Place
Lyon (France).

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
28076041
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
C CODES; DIMENSIONS; FLOW RATE; INCONEL 600; PRESSURE TUBES; PWR TYPE REACTORS; RUPTURES; STEAM GENERATORS; STRESS CORROSION
Descriptors DEC
ALLOY-NI76CR15FE8; ALLOYS; ALUMINIUM ADDITIONS; BOILERS; CHEMICAL REACTIONS; CHROMIUM ALLOYS; COMPUTER CODES; CORROSION; CORROSION RESISTANT ALLOYS; ENRICHED URANIUM REACTORS; FAILURES; HEAT RESISTANT MATERIALS; HEAT RESISTING ALLOYS; INCONEL ALLOYS; IRON ALLOYS; MATERIALS; NICKEL ALLOYS; NICKEL BASE ALLOYS; NIMONIC; POWER REACTORS; REACTORS; THERMAL REACTORS; TITANIUM ADDITIONS; TRANSITION ELEMENT ALLOYS; TUBES; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Secondary number(s)
CONF-9510432--.