Gray S. Chang

AFIP-1 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-1 was designed to demonstrate the performance of second-generation dispersion fuels at a prototypic scale with a length of 21.5 inches (54.6 cm), width of 2.25 inches (5.75 cm) and a thickness of 0.050 inch (0.13 cm). The experiment was fabricated using commercially standard practices at BWX Technology, Inc. (BWXT). The U-7Mo fuel particles were supplied by the Korean Atomic Energy Research Institute (KAERI) using equipment intended for commercial supply. Two fuel plates were tested that incorporated two different matrix compositions, Al-2Si and Al-4043.1 The following report summarizes the life of the AFIP-1 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results

AFIP-3 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-3 was designed to evaluate the performance of monolithic fuels at a prototypic scale of 2.25 inches x 21.5 inches x 0.050 inches (5.75 cm x 54.6 cm x 0.13cm). The AFIP-3 experiment was fabricated by hot isostatic pressing (HIP) and consists of two plates, one with a zirconium (Zr) diffusion barrier and one with a silicon (Si) enhanced fuel/clad interface1,2. The following report summarizes the life of the AFIP-3 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

AFIP-6 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-6 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a length prototypic to that of the ATR fuel plates (45 inches in length). The AFIP-6 test was the first test with plates in a swaged condition with longer fuel zones of approximately 22.5 inches in length1,2. The following report summarizes the life of the AFIP-6 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

GTL-1 Irradiation Summary Report

The primary objective of the Gas Test Loop (GTL-1) miniplate experiment is to confirm acceptable performance of high-density (i.e., 4.8 g-U/cm3) U3Si2/Al dispersion fuel plates clad in Al-6061 and irradiated under the relatively aggressive Booster Fast Flux Loop (BFFL) booster fuel conditions, namely a peak plate surface heat flux of 450 W/cm2. As secondary objectives, several design and fabrication variations were included in the test matrix that may have the potential to improve the high-heat flux, high-temperature performance of the base fuel plate design.1, 2 The following report summarizes the life of the GTL-1 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

AFIP-4 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

The Feasibility Study of AGR 7-Position Fuel Testing Assembly in NEFT Position

In order to support the Next Generation Nuclear Plant (NGNP) Program 2018 deployment schedule, the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program must reduce the AGR fuel irradiation testing time in the Advanced Test Reactor (ATR) from approximately 2 1/2 calendar years to 1 1/2 calendar years. The AGR fuel irradiation testing requirements are: (a) burn-up of at least 14% FIMA; (b) Fast neutron fluence (E > 0.18 MeV) – maximum 1.0 MeV) to Thermal (E 0.18 MeV) results indicate that the average fuel compact burnup and fast neutron fluence reach 14.79% FIMA and 4.16 × 1025 n/m2 , respectively. The fuel compact peak burnup reached 16.68% FIMA with corresponding fast neutron fluence for that fuel compact of 5.06 × 1025 n/m2 , which satisfied the fuel testing requirements. It is therefore concluded that accelerating the AGR fuel testing using the proposed AGR 7-position fuel test configuration in the NEFT is very feasible.Copyright