Comparison of leak opening and leak rate calculations to HDR experimental results
Creators
- 1. Gesellschaft fur Anlagen- und Reaktorsicherheit mbH, Cologne (Germany)
- 2. Kernforschungszentrum, Karlsruhe (Germany)
Description
During the last years a number of calculations of leak opening and leak rate for through cracks in piping components have been performed. Analyses are pre- or mostly post-calculations to experiments performed at the HDR facility under PWR operating conditions. Piping components under consideration were small diameter straight pipes with circumferential cracks, pipe bends with longitudinal or circumferential cracks and pipe branches with weldment cracks. The components were loaded by internal pressure and opening as well as closing bending moment. The finite element method and two-phase flow leak rate programs were used for the calculations. Results of the analyses are presented as J-integral values, crack opening displacements and areas and leak rates as well as comparisons to the experimental results
Additional details
Publishing Information
- Publisher
- American Society of Mechanical Engineers.
- Imprint Place
- New York, NY (United States)
- ISBN
- 0-7918-0993-5
- Imprint Title
- Creep, fatigue, flaw evaluation, and leak-before-break assessment
- Imprint Pagination
- 301 p.
- Journal Page Range
- p. 217-229.
Conference
- Title
- 1993 pressure vessel and piping conference.
- Dates
- 25-29 Jul 1993.
- Place
- Denver, CO (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 25025957
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- AUSTENITIC STEELS; BENDING; CRACK PROPAGATION; CRACKS; FINITE ELEMENT METHOD; FRACTURE MECHANICS; HDR REACTOR; LEAKS; PIPES; PRIMARY COOLANT CIRCUITS; TWO-PHASE FLOW; WELDED JOINTS
- Descriptors DEC
- ALLOYS; BWR TYPE REACTORS; CALCULATION METHODS; CARBON ADDITIONS; COOLING SYSTEMS; DEFORMATION; ENRICHED URANIUM REACTORS; EXPERIMENTAL REACTORS; FLUID FLOW; IRON ALLOYS; IRON BASE ALLOYS; JOINTS; MECHANICS; NUMERICAL SOLUTION; POWER REACTORS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; RESEARCH AND TEST REACTORS; STEELS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-930702--Vol.266.