Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups
The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.
- Research Organization:
- Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- AC05-76RL01830
- OSTI ID:
- 1019197
- Report Number(s):
- PNNL-SA-80796; RC0404000; TRN: US1103573
- Journal Information:
- Journal of Nuclear Materials, 412(3):294-300, Vol. 412, Issue 3; ISSN 0022-3115
- Country of Publication:
- United States
- Language:
- English
Similar Records
Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR
Evaluation of weapons grade MOX fuel performance in light water reactors using Comethe computer code