Researching | Chemical interaction layer between uranium oxide fuel pellet and zirconium alloy cladding in pressurized water reactor

Journals >NUCLEAR TECHNIQUES >Volume 46 >Issue 9 >Page 090602 > Article

NUCLEAR TECHNIQUES
Vol. 46, Issue 9, 090602 (2023)

Chemical interaction layer between uranium oxide fuel pellet and zirconium alloy cladding in pressurized water reactor

Huacai WANG^*, Dawei YANG, Huanlin CHENG, Qi TANG..., Wei WANG and Jin QIAN|Show fewer author(s)

Author Affiliations

China Institute of Atomic Energy, Beijing 102413, China

show less

DOI: 10.11889/j.0253-3219.2023.hjs.46.090602 Cite this Article

Huacai WANG, Dawei YANG, Huanlin CHENG, Qi TANG, Wei WANG, Jin QIAN. Chemical interaction layer between uranium oxide fuel pellet and zirconium alloy cladding in pressurized water reactor[J]. NUCLEAR TECHNIQUES, 2023, 46(9): 090602 Copy Citation Text

show less

References

[1] Jin H J, Kim T K. Neutron irradiation performance of Zircaloy-4 under research reactor operating conditions[J]. Annals of Nuclear Energy, 75, 309-315(2015).

[2] Ahn DH, Lim S, Lee G G et al. Study on the mechanical properties and microstructure of Zr-2.5wt%Nb pressure tube material[J]. Journal of Nuclear Materials, 523, 458-471(2019).

[3] Piro M H A, Sunderland D, Livingstone S et al. Pellet-clad interaction behavior in zirconium alloy fuel cladding[M]. Comprehensive Nuclear Materials, 248-306(2020).

[4] XU Yang, CHEN Ping, GUO Xiaoming. Design characteristics and development direction of nuclear fuel for pressurized water reactor[J]. Science and Technology Innovation Herald, 16, 62-64+67(2019).

[5] Ciszak C, Mermoux M, Miro S et al. Micro-Raman analysis of the fuel-cladding interface in a high burnup PWR fuel rod[J]. Journal of Nuclear Materials, 495, 392-404(2017).

[6] Ciszak C, Desgranges L, Garcia P et al. On the origins and the evolution of the fuel-cladding bonding phenomenon in PWR fuel rods[J]. Journal of Nuclear Materials, 520, 110-120(2019).

[7] Michel B, Nonon C, Sercombe J et al. Simulation of pellet-cladding interaction with the pleiades fuel performance software environment[J]. Nuclear Technology, 182, 124-137(2013).

[8] Riley P, Assel A. A review of microstructural features in fast reactor mixed oxide fuels[J]. Journal of Nuclear Materials, 510, 644-660(2018).

[9] Choudhuri G, Mishra P, Basu S et al. Effect of ion and neutron irradiation on oxide of PHWR fuel tube material[J]. Journal of Nuclear Materials, 514, 12-27(2019).

[10] WANG Huacai, CHENG Huanlin, SONG Wulin et al. Raman characteristics analysis of the fuel-cladding chemical interaction layer for intact and leak PWR fuel rods[J]. Atomic Energy Science and Technology, 57, 619-629(2023).

[11] Guerain M, Duriez C, Grosseau-Poussard J L et al. Review of stress fields in Zirconium alloys corrosion scales[J]. Corrosion Science, 95, 11-21(2015).

[12] Platt P, Polatidis E, Frankel P et al. A study into stress relaxation in oxides formed on zirconium alloys[J]. Journal of Nuclear Materials, 456, 415-425(2015).

[13] Kurpaska L, Favergeon J, Grosseau-Poussard J L et al. In-situ stress analysis of the Zr/ZrO2 system as studied by Raman spectroscopy and deflection test in monofacial oxidation techniques[J]. Applied Surface Science, 385, 106-112(2016).

[14] Kurpaska L, Favergeon J, Lahoche L et al. Zirconia layer formed by high temperature oxidation of pure zirconium: stress generated at the zirconium/zirconia interface[J]. Oxidation of Metals, 79, 261-277(2013).

[15] Park Y S, Kim J G, Kwon H M et al. Microstructural changes of fuel claddings for pressurized water reactors fuel after irradiation[J]. Asian Journal of Chemistry, 25, 7009-7012(2013).

[16] Tanaka K, Maeda K, Sasaki S et al. Fuel-cladding chemical interaction in MOX fuel rods irradiated to high burnup in an advanced thermal reactor[J]. Journal of Nuclear Materials, 357, 58-68(2006).

[17] Kim K T. UO2/Zry-4 chemical interaction layers for intact and leak PWR fuel rods[J]. Journal of Nuclear Materials, 404, 128-137(2010).

[18] Ciszak C, Mermoux M, Miro S et al. Micro-Raman analysis of the fuel-cladding interface in a high burnup PWR fuel rod[J]. Journal of Nuclear Materials, 495, 392-404(2017).

[19] Garner A, Gholinia A, Frankel P et al. The microstructure and microtexture of zirconium oxide films studied by transmission electron backscatter diffraction and automated crystal orientation mapping with transmission electron microscopy[J]. Acta Materialia, 80, 159-171(2014).

[20] Bouvier P, Godlewski J, Lucazeau G. A Raman study of the nanocrystallite size effect on the pressure-temperature phase diagram of zirconia grown by zirconium-based alloys oxidation[J]. Journal of Nuclear Materials, 300, 118-126(2002).

[21] Nogita K, Une K. Formation of pellet-cladding bonding layer in high burnup BWR fuels[J]. Journal of Nuclear Science and Technology, 34, 679-686(1997).

[22] Lach T G, Edwards D J, Buck E C et al. Fission recoil-induced microstructural evolution of the fuel-cladding interface [FCI] in high burnup BWR fuel[J]. Journal of Nuclear Materials, 521, 120-125(2019).

Get PDF(in Chinese)

Figures&Tables (8)

References (22)

Copy Citation Text

Huacai WANG, Dawei YANG, Huanlin CHENG, Qi TANG, Wei WANG, Jin QIAN. Chemical interaction layer between uranium oxide fuel pellet and zirconium alloy cladding in pressurized water reactor[J]. NUCLEAR TECHNIQUES, 2023, 46(9): 090602

Download Citation

Save the article for my favorites

Paper Information

Recommended Topics

laser devices and laser physics

Lasers and Laser Optics

laser manufacturing

Instrumentation, Measurement and Metrology

Set citation alerts for the article

Please enter your email address