Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments

Technical Report · Sun Sep 30 04:00:00 EDT 2018

DOI:https://doi.org/10.2172/1498255· OSTI ID:1498255

Harp, Jason M ^[1]; Capriotti, Luca ^[1]; Cappia, Fabiola ^[1]

---

1. Idaho National Lab. (INL), Idaho Falls, ID (United States)

A long term research focus of the US Department of Energy (DOE) Advanced Fuels Campaign (AFC) is to investigate technologies that can allow for increased actinide utilization in nuclear fuel for fast neutron spectrum reactors. This includes both the incorporation of minor actinides into nuclear fuel for transmutation and high utilization of actinides through “ultra high” burnup fuel. This goal is defined as fuel that can reach a burnup of 30-40 percent fission per initial metal atom (%FIMA) which is higher than what has historically been achieved [1],[2]. Historically, Experimental Breeder Reactor II (EBR-II) Mark III/IIIA/IV driver fuel (U 10Zr fuel with stainless steel 316 or ferritic martensitic steel HT-9 cladding) was qualified to 10%FIMA, but many experimental assemblies and fuel pins achieved higher peak burnup up to near 20%FIMA without failure [3]. The motivation of the irradiation tests explored in this work was to screen candidate alloys and forms that could tolerate very high burnup irradiations of up to 30 % FIMA. In these irradiations, termed AFC 3C, AFC-3D and AFC-4A, uranium-based alloys and forms were irradiated in a prototypic fast reactor spectrum to investigate fuel performance. Irradiations were performed in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). The performance of these fuels is compared to the historic performance of U-10Zr [4]. Metallic fuel for fast reactors has a long history that has been reviewed several time in the literature [3]–[8].

• Research Organization: Idaho National Laboratory (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1498255
• Report Number(s): INL/EXT--18-51447-Rev000
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
AFC-3C, AFC-3D, and AFC-4A experiments
Basline Postirradiation Examination