Probabilistic structural integrity assessment for Inconel690 alloy steam generator tube with volume defect
Creators
- 1. School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China)
- 2. Suzhou Nuclear Power Research Institute, Suzhou 215000 (China)
Description
Highlights: • Non-destructive examination and 350℃ bursting pressure tests of defected tube. • Probabilistic bursting pressure model of Inconel690 tube with volume defect. • Fit for service criterion for Inconel690 tube with less conservatism. Fit for service(FFS) assessment of Steam Generator Tube(SGT) plays a key role in safety operating of Pressurized Water Reactors. Engineering practices have emerged probabilistic method is a powerful tool in reducing redundant safety margin of FFS assessment strategy. The aim of this work is to develop a probabilistic structural integrity assessment method for typical Inconel690 alloy SGT with volume defect. The study consists of four parts: Firstly, an experiment program including tensile test, eddy current examination and SGT bursting test were introduced. Then, potential bursting pressure models were analyzed based on test dataset. Proper bursting pressure model was suggested. Thirdly, probabilistic bursting pressure prediction model was developed based on Monte Carlo Simulation. At last, parametric study was performed and an application instance on anti-vibration bar fretting wear was presented. The results indicated the proposed SGT structural integrity assessment method is less conservative than current repair criterions. This study represents a further step towards developing of smart FFS assessment of Nuclear Power Plant.
Availability note (English)
Available from http://dx.doi.org/10.1016/j.nucengdes.2020.110949Additional details
Identifiers
- DOI
- 10.1016/j.nucengdes.2020.110949;
- PII
- S002954932030443X;
Publishing Information
- Journal Title
- Nuclear Engineering and Design
- Journal Volume
- 371
- Journal Page Range
- vp.
- ISSN
- 0029-5493
- CODEN
- NEDEAU
INIS
- Country of Publication
- Netherlands
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 54014908
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS; S36: MATERIALS SCIENCE;
- Descriptors DEI
- ALLOYS; COMPUTERIZED SIMULATION; EDDY CURRENTS; MONTE CARLO METHOD; NUCLEAR POWER PLANTS; PARAMETRIC ANALYSIS; PROBABILISTIC ESTIMATION; PWR TYPE REACTORS; SAFETY MARGINS; STEAM GENERATOR TUBE RUPTURE; STEAM GENERATORS
- Descriptors DEC
- ACCIDENTS; BOILERS; CALCULATION METHODS; CURRENTS; ELECTRIC CURRENTS; ENRICHED URANIUM REACTORS; NUCLEAR FACILITIES; POWER PLANTS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SIMULATION; THERMAL POWER PLANTS; THERMAL REACTORS; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Copyright
- Copyright (c) 2020 Elsevier B.V. All rights reserved.