Research Papers

Fission Product Release Under Supercritical Water-Cooled Reactor Conditions

[+] Author and Article Information
D. Guzonas

Canadian Nuclear Laboratories, Chalk River Laboratories, CNL,
Chalk River, ON K0J 1J0, Canada
e-mail: david.guzonas@cnl.ca

L. Qiu

Canadian Nuclear Laboratories, Chalk River Laboratories, CNL,
Chalk River, ON K0J 1J0, Canada
e-mail: liyan.qiu@cnl.ca

S. Livingstone

Canadian Nuclear Laboratories, Chalk River Laboratories, CNL,
Chalk River, ON K0J 1J0, Canada
e-mail: steve.livingstone@cnl.ca

S. Rousseau

Canadian Nuclear Laboratories, Chalk River Laboratories, CNL,
Chalk River, ON K0J 1J0, Canada
e-mail: stephane.rousseau@cnl.ca

Manuscript received May 22, 2015; final manuscript received July 31, 2015; published online February 29, 2016. Assoc. Editor: Thomas Schulenberg.

ASME J of Nuclear Rad Sci 2(2), 021010 (Feb 29, 2016) (6 pages) Paper No: 15-1092; doi: 10.1115/1.4031381 History: Received May 22, 2015; Accepted July 31, 2015

Most supercritical water-cooled reactor (SCWR) concepts being considered as part of the Generation IV initiative are direct cycle. In the event of a fuel defect, the coolant will contact the fuel pellet, potentially releasing fission products and actinides into the coolant and transporting them to the turbines. At the high pressure (25 MPa) in an SCWR, the coolant does not undergo a phase change as it passes through the critical temperature in the core, and nongaseous species may be transported out of the core and deposited on out-of-core components, leading to increased worker dose. It is therefore important to identify species with a high risk of release and develop models of their transport and deposition behavior. This paper presents the results of preliminary leaching tests in SCW of U-Th simulated fuel pellets prepared from natural U and Th containing representative concentrations of the (inactive) oxides of fission products corresponding to a fuel burnup of 60  GWd/ton. The results show that Sr and Ba are released at relatively high concentrations at 400°C and 500°C.

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Grahic Jump Location
Fig. 1

Noble gas and iodine release versus decay constant (λ) from EVESR superheat fuel rod KB-40 (Adapted from Ref. [8]). Y-axis is release rate divided by the yield (y) and the decay constant. Data from April 2, 1966: reactor power 13 MW(t) and outlet steam temperature 847°F. The peak cladding temperature at the time of failure was estimated to be 777°C

Grahic Jump Location
Fig. 2

Weight change for the SIMFUEL samples tested at 400°C and 25 MPa at different exposure times. Samples 1, 2, and 3 were exposed for 1 day, samples 4, 5, and 6 for 3 days, and samples 7, 8, and 9 for 7 days

Grahic Jump Location
Fig. 3

Concentrations of autoclave corrosion products in the test solution as a function of exposure time at 400°C and 25 MPa

Grahic Jump Location
Fig. 4

Concentrations of Sr and Ba in the test solution as a function of exposure time at 400°C and 500°C, 25 MPa

Grahic Jump Location
Fig. 5

Solubility of UO2(c) at 500°C as a function of pH using a Ni-NiO buffer. Adapted from Red’kin et al. [27]

Grahic Jump Location
Fig. 6

Thermodynamically calculated solubility of ThO2 at 25°C as a function of pH




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