References

SCIANTIX papers

Pizzocri D. et al (2020). SCIANTIX: A new open source multi-scale code for fission gas behaviour modelling designed for nuclear fuel performance codes. Journal of Nuclear Materials, 532, 152042.

Zullo G. et al (2023). The SCIANTIX code for fission gas behaviour: Status, upgrades, separate-effect validation, and future developments. Journal of Nuclear Materials, 587, 154744.

Validation

Zullo G. et al (2024). Integral-scale validation of the SCIANTIX code for Light Water Reactor fuel rods Journal of Nuclear Materials, 601, 155305.

Models

Barani T. et al (2020). Modeling high burnup structure in oxide fuels for application to fuel performance codes. part I: High burnup structure formation. Journal of Nuclear Materials, 539, 152296.

Barani T. et al (2022). Modeling high burnup structure in oxide fuels for application to fuel performance codes. Part II: Porosity evolution. Journal of Nuclear Materials, 563, 153627.

Blackburn P.E. (1973). Oxygen pressures over fast breeder reactor fuel (I) A model for UO2+-x. Journal of Nuclear Materials, 46, 244-252.

Cechet A. et al (2021). A new burn-up module for application in fuel performance calculations targeting the helium production rate in (U,Pu)O2 for fast reactors. Nuclear Engineering and Technology, 53, 1893-1908.

Claisse A. and Van Uffelen P., (2015). Journal of Nuclear Materials, 466, 351-356.

Cognini L. et al (2021). Towards a physics-based description of intra-granular helium behaviour in oxide fuel for application in fuel performance codes. Nuclear Engineering and Technology, 53, 562-571.

Khvostov G. et al (2005). Approaches to Modeling of High Burn-up Structure and Analysis of its Effects on the Behaviour of Light Water Reactor Fuels in the START-3 Fuel Performance Code. WRFPM-2005, Kyoto, Japan.

Lewis B.J. et al (1995). Modelling the release behaviour of cesium during severe fuel degradation. Journal of Nuclear Materials, 227, 83-109.

Lösönen P. (2002). Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel. Journal of Nuclear Materials, 304, 29-49.

Pastore G. et al (2013). Physics-based modelling of fission gas swelling and release in UO2 applied to integral fuel rod analysis. Nuclear Engineering and Design, 256, 75-86.

Pizzocri D. et al (2018). Physics-based modelling of fission gas swelling and release in UO2 applied to integral fuel rod analysis. Journal of Nuclear Materials, 502, 323-330.

Turnbull J.A., Beyer C.E. (2010). Background and Derivation of ANS-5.4 Standard Fission Product Release Model.

White R.M. and Tucker P.G. (1983). A new fission-gas release model. Journal of Nuclear Materials, 118, 1-38.

Van Uffelen P. (2002). PhD thesis, SCK•CEN Reports; No. BLG-907, SCK CEN.

White R.M. (2004). The development of grain-face porosity in irradiated oxide fuel. Journal of Nuclear Materials, 325, 61-77.

Zullo G. et al (2024). Two-phase modelling for fission gas sweeping in restructuring nuclear oxide fuel. Nuclear Enginnering and Design, 429, 113602.

Experiments

Abrefah H. et al (1994). High temperature oxidation of UO2 in steam-hydrogen mixtures. Journal of Nuclear Materials, 208, 98-110.

Ainscough J.B. et al (1973). Isothermal grain growth kinetics in sintered UO2 pellets. Journal of Nuclear Materials, 49, 117-128.

Carter G. and Lay G.R. (1970). Surface-controlled oxidation-reduction of UO2. Journal of Nuclear Materials, 36, 77-86.

Talip Z. et al (2014). Thermal diffusion of helium in 238Pu-doped UO2. Journal of Nuclear Materials, 445, 117-127.

Van Uffelen P. et al (2013). An experimental study of grain growth in mixed oxide samples with various microstructures and plutonium concentrations. Journal of Nuclear Materials, 434, 287-290.

Correlations

Cognini L. et al (2018). Helium solubility in oxide nuclear fuel: Derivation of new correlations for Henry’s constant. Nuclear Engineering and Design, 340, 240-244.

Luzzi L. et al (2018). Helium diffusivity in oxide nuclear fuel: Critical data analysis and new correlations. Nuclear Engineering and Design, 330, 265-271.

Reynolds A.B. and Burton C. (1979). Grain-boundary diffusion in uranium dioxide: The correlation between sintering and creep and a reinterpretation of creep mechanism. Journal of Nuclear Materials, 82, 22-25.

Bittel R. et al (1969). Steam Oxidation Kinetics and Oxygen Diffusion in UOz at High Temperatures. Journal of the American Ceramic Society, 52, 446-451.

Reviews

Matzke H. (1980). Gas release mechanisms in UO2—a critical review. Radiation Effects, 53, 219-242.

Algorithms

Zullo G. et al (2022). On the use of spectral algorithms for the prediction of short-lived volatile fission product release: Methodology for bounding numerical error. Nuclear Engineering and Technology, 54, 1195-1205.

Other

Cox B. et al (1986). Oxidation of UO2 in air and steam with relevance to fission product releases. NUREG/CP-0078, U.S. NRC.

Evans J.H. (1994). Bubble diffusion to grain boundaries in UO2 and metals during annealing: a new approach. Journal of Nuclear Materials, 210, 21-29.

Ham F.S. (1958). Theory of diffusion-limited precipitation. Journal of Physics and Chemistry of Solids, 6, 335-351.

Imamura M. and Une K. (1997). High temperature steam oxidation of U02 fuel pellets. Journal of Nuclear Materials, 247, 131-137.

Massih A.R. (2018). UO2 Fuel Oxidation and Fission Gas Release, Swedish Radiation Safety Authority (SSM).

Morel B., Serose N., Gamaury S., Ph. Dehaudt, “Cinetique d’Oxydation du Combustible par la Vapeur d’Eau; Influence du S/V,” Commissariat a l’Energie Atomique, Report NT/DTP/SECC no. DR94-55, June 1994..

Olander D.R. and Wongsawaeng D. (2006). Re-solution of fission gas – A review: Part I. Intragranular bubbles. Journal of Nuclear Materials, 354, 94-109.

Ronchi C. (2007). Thermophysical properties affecting safety and performance of nuclear fuel. High Temperatures, 45, 552-571.

Turnbull J.A. (1971). The re-solution of fission-gas atoms from bubbles during the irradiation of UO2 at an elevated temperature. Journal of Nuclear Materials, 38, 203.

Turnbull J.A. et al (1988). The diffusion coefficient for fission gas atoms in uranium dioxide. IWGFPT-32, Preston, UK, Sep 18-22.