Irradiation Issues and Material Selection for Canadian SCWR Components

[+] Author and Article Information
L. Walters

Canadian Nuclear Laboratories,
286 Plant Road,
Chalk River, ON K0J 1J0, Canada
e-mail: lori.walters@cnl.ca

M. Wright, D. Guzonas

Canadian Nuclear Laboratories,
286 Plant Road,
Chalk River, ON K0J 1J0, Canada
e-mail: michael.wright@cnl.ca

1Corresponding author.

2This is the inlet channel for the coolant. At the bottom of the channel the coolant reverses direction and flows back upward, over the fuel, within the annulus between the central flow tube and the liner tube protecting the insulator and outer pressure tube.

3Note that there is also an outer liner between the insulator and the pressure tube and this is not exposed to the SCW coolant.

Manuscript received April 30, 2017; final manuscript received October 3, 2017; published online May 16, 2018. Assoc. Editor: Thomas Schulenberg.

ASME J of Nuclear Rad Sci 4(3), 031005 (May 16, 2018) (10 pages) Paper No: NERS-17-1045; doi: 10.1115/1.4038367 History: Received April 30, 2017; Revised October 03, 2017

The Canadian super critical water-cooled reactor (SCWR) concept requires materials to operate at higher temperatures than current generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux, and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in-core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking, and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.

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Grahic Jump Location
Fig. 1

Details showing the bottom of a fuel channel with the fuel assembly

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Fig. 2

Canadian SCWR fuel channel with fuel assembly

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Fig. 3

Neutron energy spectra in the Canadian SCWR at a high flux location

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Fig. 4

Canadian SCWR neutron flux spectrum compared to 59Ni (n, p) and (n, α) reaction cross sections as a function of neutron energy

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Fig. 5

Neutron energy spectra of irradiation sites in HFIR and ATR test reactors compared to that of the Canadian SCWR

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Fig. 6

Comparison of damage versus years for the SCWR pressure tube and Excel materials in the HFIR and ATR reactors

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Fig. 7

Comparison of the neutron flux spectrum in HFIR, EBR-II, and Canadian SCWR

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Fig. 8

Helium/dpa for Alloy 800H in HFIR and Canadian SCWR neutron spectra

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Fig. 9

Comparison of helium produced in Alloy 800H in Canadian SCWR, HFIR, and EBR-II

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Fig. 10

Comparison of dpa in Alloy 800H produced in Canadian SCWR, HFIR, and EBR-II neutron spectra



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