Preliminary Analysis of the Cladding Mechanical Behavior of a Nuclear Superheat Boiling Water Reactor

Document Type: Original Article

Authors

Department of Nuclear Engineering, Science and Research branch, Islamic Azad University, Tehran, Iran

Abstract

In the present study, investigation of mechanical behaviour of the fuel cladding material for a nuclear superheat Boiling Water Reactor with annular fuel rods, is carried out. In this design, each annular fuel element is cooled internally by steam and externally by water. For the fuel cladding material, radiation embitterment and irradiation-assisted stress corrosion cracking (IASCC) are the most important issues that have to be taken into account. Hence, for cladding, two materials are considered. Preliminary thermal expansion and stress analysis have been done for a fresh (begin of cycle) ASBWR (Annular-fuelled Superheat Boiling Water Reactor) fuel element. The purpose of these analysis is to investigate the stress distribution and thermal expansion of the cladding in the initial phase of operation. The results show that there is a noticeable difference in the axial expansion between the inner and outer claddings. For T91 (modified 9Cr-1Mo steel) cladding, the maximum axial thermal growth of the inner cladding is 22.12 mm, which is about 9.7 mm more than the outer cladding. For Inconel 718 cladding, the results are     27.8 mm and 13.4 mm, respectively.

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References

[1]     Novick, M., et al., Developments in Nuclear Superheat, Proceedings of the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, Aug. 31, Sep. 9, 1964.

[2]     Chih Ko, Yu., Conceptual Design of an Annular-Fueled Superheat Boiling Water Reactor, MIT Ph.D Thesis, Jun. 2010.

[3]     Sapundjiev, D., Al Mazouzi, A., and Van Dyck, S., A Study of the Neutron Irradiation Effects on the Susceptibility to Embrittlement of A316L and T91 Steels in Lead-Bismuth Eutectic, Journal of Nuclear Materials, Vol. 356, No. 1-3, Sep. 2006, pp. 229-236.

[4]     Chen, ., Sridharan, K., and Allen, T., Corrosion Behavior fferritic-Martensitic Steel T91 in Supercritical Water, Corrosion Science, Vol. 48, No. 9, 2006, pp. 2843-2854.   

[5]     Fazzolae, R., Gas-Cooled Fast Breeder Reactor Designs with Advanced Fuel and Cladding, Nuclear Engineering and Design, Vol. 40, No. 1, Jan. 1977, pp. 191-201.

[6]     Incone 718 Properties, On-line Resource, Available from World Wide Web: http://www.espimetals.com/tech/inconel7l8.pdf http://www.espimetals.com/technicaldata.htm

[7]     Ashrafi Nik, M., Dec. Thermo Hydraulic Optimization of the EURISOL DS Target, No. 2, TM-34-06-04, 2006.

[8]     Harvey, J. F., Theory and Design of Pressure Vessels, Springer, Sep. 1991.

ADMT Journal

Volume 11, Issue 2
Spring 2018
Pages 39-45

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