Performance of FCCI barrier foils for U–Zr–X metallic fuel

Ryu, Ho Jin; Lee, Byoung Oon; Oh, Seok Jin; Kim, Junhwan; Lee, Chan Bock

doi:10.1016/j.jnucmat.2009.03.011

Cited by 65 publications

(20 citation statements)

References 22 publications

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“…9 and 10 show an axial and radial view respectively of the reference B&B fuel rod design using porous-plug fission gas venting and annular metallic fuel. The annular fuel is mechanically bonded to a 30 lm vanadium (or possible Zirconium) liner (Ryu et al, 2009), which acts as a diffusion barrier to avoid fuel/cladding chemical interaction (FCCI).…”

Section: Suggested Bandb Core Fuel Rod Designmentioning

confidence: 99%

Design and performance of 2D and 3D-shuffled breed-and-burn cores

Qvist

Hou

Greenspan

2015

Annals of Nuclear Energy

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Section: Suggested Bandb Core Fuel Rod Designmentioning

confidence: 99%

Design and performance of 2D and 3D-shuffled breed-and-burn cores

Qvist

Hou

Greenspan

2015

Annals of Nuclear Energy

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“…Additionally, formation of an interaction layer can reduce cladding thickness and decreases its load-bearing capability, consequently decreasing the life of a fuel-bearing element. [3] Understanding fuel constituent redistribution that alters alloy composition and affects fuel performance is imperative for safe operation of the reactor. However, to comprehend irradiation-induced changes in U-Pu-Zr alloys and consequent FCCI, the phases and microstructure of unirradiated fuels should be characterized first.…”

Section: Introductionmentioning

confidence: 99%

Fuel-Cladding Interaction Between U-Pu-Zr Fuel and Fe

Aitkaliyeva

Madden

Miller

et al. 2015

Metallurgical and Materials Transactions E

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This work investigates fuel-cladding chemical interaction (FCCI) between U-25Pu-14Zr (in wt pct) fuel and pure Fe at elevated temperatures, understanding of which is critical for evaluation of the fuel performance. Phases and microstructure formed in the quaternary uranium-plutonium-zirconium-iron (U-Pu-Zr-Fe) system were characterized using the transmission electron microscopy technique. Phases formed within the FCCI layer were identified using selective area electron diffraction (SAED) analysis as Fe 2 U (Fd-m), Fe 2 Zr (Fd-3m), a-U (Cmcm), Fe 2 Pu (Fd-3m), b-Pu (C12/m1), and b-Zr (Im-3m).

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“…In the case of the U-10Zr/T92 specimen, the migration phenomena of U, Zr, Fe, and Cr as well as the Nd lanthanide fission products were observed at the eutectic melting region. The measured penetration 5.6 at.% (U-19Pu-10Zr) [2] 11.1 at.% (U-19Pu-10Zr) [2] 2.44 at.% (U-10Zr): in this work Unirradiated U-10Zr [18][19][20]   …”

Section: Discussionmentioning

confidence: 49%

“…These results are presented in Figure 7 along with the experimental results of the penetration rate with 1 h heating data in the FBTA tests which were reported in the literature [2,3]. In addition, the measured eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (ferritic martensitic steel, HT9 or Gr.91) cladding specimens through the out-of-pile heating test which are available in the literature [18][19][20] are also presented in Figure 7. In the case of the heating test for unirradiated fuel with HT9 at 700 • C during 96 h, the measured penetration depth along the cladding direction was approximately 8 µm (penetration rate = 2.3 × 10 −5 µms −1 Figure 13, the measured eutectic penetration rates for the irradiated fuel specimens are higher than those for the unirradiated U-10Zr specimen.…”

Section: Results Of Irradiated Fuel After High-temperature Heating Testmentioning

confidence: 76%

Results of High-Temperature Heating Test for Irradiated U-10Zr(-5Ce) with T92 Cladding Fuel

Kim¹,

Cheon²,

Kim³

et al. 2016

Metals

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A microstructure observation using an optical microscope, SEM and EPMA was performed for the irradiated U-10Zr and U-10Zr-5Ce fuel slugs with a T92 cladding specimen after a high-temperature heating test. Also, the measured eutectic penetration rate was compared with the value predicted by the existing eutectic penetration correlation being used for design and modeling purposes. The heating temperature and duration time for the U-10Zr/T92 specimen were 750 • C and 1 h, and those for the U-10Zr-5Ce/T92 specimen were 800 • C and 1 h. In the case of the U-10Zr/T92 specimen, the migration phenomena of U, Zr, Fe, and Cr as well as the Nd lanthanide fission product were observed at the eutectic melting region. The measured penetration rate was similar to the value predicted by the existing eutectic penetration rate correlation. In addition, when comparing with measured eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (ferritic martensitic steel, HT9 or Gr.91) cladding specimens which had been reported in the literature, the measured eutectic penetration rate for the irradiated fuel specimen was higher than that for the unirradiated U-10Zr specimen. In the case of the U-10Zr-5Ce/T92 specimen in which there had been a gap between the fuel slug and cladding after the irradiation test, the eutectic melting region was not found because contact between the fuel slug and cladding did not take place during the heating test.

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Performance of FCCI barrier foils for U–Zr–X metallic fuel

Cited by 65 publications

References 22 publications

Design and performance of 2D and 3D-shuffled breed-and-burn cores

Design and performance of 2D and 3D-shuffled breed-and-burn cores

Fuel-Cladding Interaction Between U-Pu-Zr Fuel and Fe

Results of High-Temperature Heating Test for Irradiated U-10Zr(-5Ce) with T92 Cladding Fuel

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