[en] This study presents experimentally simulated radiation-damage-induced microstructural evolution of FC-92B with corresponding computational simulation results obtained using the Stopping and Range of Ions in Matter (SRIM). Prior to the full-scale assessment of radiation performance of FC-92, an F/M steel developed as PGSFR fuel cladding material, preliminary investigation on high burnup microstructural evolution of the cladding material was conducted utilizing ion-beam irradiation to be complementarily analyzed with in-pile test results. The alloy irradiated using 60 keV

[en] To reuse fissile materials through fuel cycle in future nuclear energy system development and achieve safety, economics, optimization of spent fuel management, an information of isotopic fissile content(U235, Pu239, Pu241, MA) is required to be provided basically. LSDS technology development was done to assay isotopic fissile contents. LSDS system consists of spectrometer, source neutron, measurement, data process. In 1st phase (2012-2013), key measurement technology, neutron source technology, conceptual design of detection system and shielding design were done. In the spectrometer, neutron spectrum and resolution analysis, detector design and response, detection model were performed. The optimum shielding model was proposed by shielding dose simulation. In 2nd phase (2014-2016), key parameters were determined in optimized spectrometer, fission detection sensitivity, accelerator and beam dump design, target design and neutron yield analysis, cooling system, material activation, optimized shielding model were done, and finally, mathematical assay model was proposed with software development to isotopic fissile content. Additionally, correction methodology was established by fission signal analysis. Using real U235(4.8%) and Pu239(47g, 91g), fission data was obtained. In fissile assay using various fissile materials, the content of uranium and plutonium was assay with 1-3% error. In the unknown sample, the content of plutonium was analyzed in 1-2% uncertainty. The isotopic fissile content assay technology is required in the reuse by fuel cycle for success of future nuclear energy system development. Moreover, such an advanced new fissile assay technology will lead the international nuclear area and contribute to technology export.

[en] Highlights: • Void swelling was studied in self-ion irradiated HT.9, Gr.92, FC92-B, and FC92-N. • FC92-N showed the greatest swelling resistance of 1.76% at 240 dpa and 2.17% 318 dpa. • M₂X evolved only in FC92 series, in which bimodal swelling profiles were observed. • RIP and outward Cr sinking to the surface synergistically formed a low-alloyed zone. • RIP modified swelling-depth profiles, and in turn, determined double peak swelling. The radiation responses of two newly developed ferritic/martensitic steels, FC92-B and -N, were tested in comparison to reference alloys HT9 and Gr.92. Ion irradiations on the steels were performed up to 480 dpa at 475°C using 3.5-MeV Fe⁺⁺ ions with a helium pre-implantation of 1 appm/dpa. Void swelling and M₂X precipitation were characterized using FE-TEM and EDS. Swelling resistance was the greatest in FC92-N, which showed suppressed void nucleation and growth. The swelling rate in FC92-N was determined as 0.005 %/dpa, indicating that FC92-N did not reach the steady-state swelling regime with void nucleation behavior. The least swelling-resistant alloy was HT9 with a swelling rate of 0.048 %/dpa. Cr-rich carbide, M₂X, was observed in only 9Cr-FC92 series; however, its formation did not depend on radiation damage. This exceptional M₂X evolution in FC92 series may be attributed to B and N alloying, which resulted in suppressed M₂₃C₆ carbide formation during metallurgical production and sequentially high C contents in the alloy solution of FC92 series. A narrower range (800 nm) of M₂X evolution compared to that of cavity formation (1,000 nm) indicates that radiation-induced precipitation (RIP) is more sensitive to the injected interstitial effect. Precipitation-induced Cr depletion and preferential interstitial outward sinking to the free surface synergistically modified local chemical composition before void evolution and led to double-peak swelling by locally forming a low-alloyed zone. This study provides the first experimental evidence that RIP modifies the swelling–depth profiles and in turn, determines double-peak swelling in ion-irradiated steels.

[en] Thermal conductivity of spark plasma sintered UN/Gd₂O₃ and UN/GdN composite pellets with various compositions of Gd₂O₃ (5.0 to 15.0 wt%) and GdN (3.5 to 38.4 wt%) was measured from 25 °C to 1000 °C using laser flash analyzer (LFA). Overall thermal conductivity of both composite pellets decreases with increasing Gd compositions for the entire temperature range measured. The UN/GdN pellets sintered at high temperature (2000 °C) contained solid-solution (U_1-x,Gd_x)N phase and exhibited 20–65 % higher thermal conductivity than the UN/Gd₂O₃ pellets. We also confirmed a two-phase structure in low temperature (1800 °C) sintered UN/GdN pellets, which exhibited 14–25 % higher thermal conductivity than solid-solution UN/GdN pellets. This result may indicate that relatively higher thermal conductivity of UN/GdN composite as burnable absorber fuel could be further enhanced with an optimized microstructure.

[en] Accelerated corrosion tests of Al-B₄C neutron absorber, equivalent to 37 months in a spent nuclear fuel pool, were conducted on three different period-installed surveillance coupons (33, 52, and 92 months) to further investigate the underlying mechanisms of premature surface corrosion and ¹⁰B depletion which we had recently reported in another study. Microstructure characterization, electrochemical analysis, and neutron attenuation tests were conducted after the corrosion tests, and two types of galvanic corrosion, Al matrix/stainless steel and Al matrix/B₄C particles, were discovered. The duplex oxide layer comprised of amorphous oxide and Al(OH)₃ films and pit corrosion were also observed on the surface with reduced densities (10B areal density.

[en] Highlights: • Phase transformation temperatures and enthalpies of U–0.1, 2, 5, 10, 20, 30, 40, and 50 wt% Zr alloys were measured using DSC–TGA. • The phase transformation of the (α-U, γ_2) phase to the (β-U, γ_2) phase at ∼662 °C was not evident in Zr-rich (>10 wt%) U–Zr alloys. • The absence of the phase transformation is rather consistent with the older U–Zr phase diagram that was experimentally assessed in the 1950s. • The current U–Zr binary alloy phase diagram may need to be revisited regarding the determination of the range of the (β, γ_2) phase zone. - Abstract: The solid phase transformation behavior of uranium–zirconium (U–Zr) alloys (U–0.1, 2, 5, 10, 20, 30, 40, and 50 wt% Zr) was observed using differential scanning calorimetry (DSC) with thermogravimetric analysis (TGA). The phase transformation temperatures and enthalpies were measured from the alloys annealed at 600 °C for 72, 168, and 672 h. The observations indicated distinctive mismatches between the measured data and the existing U–Zr alloy phase diagram. Most notably, the phase transformation of the (α-U, γ_2) phase to the (β-U, γ_2) phase at ∼662 °C was not evident in Zr-rich (> 10 wt%) U–Zr alloys, while only two phase transformations were evident in the U–10Zr and U–20Zr alloys compared to the three isotherm lines extended over the two compositions in the current phase diagram. The absence of the phase transformation is rather consistent with the older U–Zr phase diagram that was experimentally assessed in the 1950s. This observation may lead to the conclusion that the (β-U, γ_2) phase region is not correctly represented in the Zr-rich portion, or the hyper-monotectoid region, of the current U–Zr alloy phase diagram. It is evident that the phase diagram needs to be experimentally revisited to provide more reliable information for the development of metallic nuclear fuel performance models, if such models are to include phase-relevant effects, such as fuel constituent redistribution and fission gas swelling

[en] An investigation of the decomposition of the high temperature γ phase in as-cast and quenched U–Zr alloys was conducted. Differential scanning calorimetry data clearly showed δ⇌γ transformations in alloys with <10 wt% Zr while XRD data did not contain any peaks which uniquely identify it's presence. Since δ phase forms via ω transformation, a comparison of the theoretical diffraction patterns for ω and δ revealed that the intensities of the peaks which uniquely identify the existence of δ when α-U is present, were either very weak, or were zero in ω, suggesting that the ambiguity can be explained if the phase present in these alloys is ω as opposed to δ. Our data are consistent with the presence of δ and ω in as-cast and quenched U–50Zr alloy, respectively, and (α + ω) in rest of the as-cast and quenched alloys. Based on the experimental data, the transformation sequence from γ phase in U–Zr alloys is proposed.