Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis
- Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (Japan)
- Quantum Beam Science Directorate, Japan Atomic Energy Agency (Japan)
- Dept. of Nuclear Engineering and Management, Univ. of Tokyo (Japan)
A double-wall-tube is nominated for the steam generator heat transfer tube of future sodium fast reactors (SFRs) in Japan, to decrease the possibility of sodium/water reaction. The double-wall-tube consists of an inner tube and an outer tube, and they are mechanically contacted to keep the heat transfer of the interface between the inner and outer tubes by their residual stress. During long term SG operation, the contact stress at the interface gradually falls down due to stress relaxation. This phenomenon might increase the thermal resistance of the interface and degrade the tube heat transfer performance. The contact stress relaxation can be predicted by numerical analysis, and the analysis requires the data of the initial residual stress distributions in the tubes. However, unclear initial residual stress distributions prevent precious relaxation evaluation. In order to resolve this issue, a neutron diffraction method was employed to reveal the tri-axial (radius, hoop and longitudinal) initial residual stress distributions in the double-wall-tube. Strain gauges also were used to evaluate the contact stress. The measurement results were analyzed using a JAEA's structural computer code to determine the initial residual stress distributions. Based on the stress distributions, the structural computer code has predicted the transition of the relaxation and the decrease of the contact stress. The radial and longitudinal temperature distributions in the tubes were input to the structural analysis model. Since the radial thermal expansion difference between the inner (colder) and outer (hotter) tube reduces the contact stress and the tube inside steam pressure contributes to increasing it, the analytical model also took these effects into consideration. It has been conduced that the inner and outer tubes are contacted with sufficient stresses during the plant life time, and that effective heat transfer degradation dose not occur in the double-wall-tube SG. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22105987
- Resource Relation:
- Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 6 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
- Country of Publication:
- United States
- Language:
- English
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