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Mohamad, A. B.; 山下 真一郎; 根本 義之; 阿部 陽介; Pham, V. H.; 井岡 郁夫; 相馬 康孝; 石島 暖大; 佐藤 智徳; Rizaal, M.; et al.
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 8 Pages, 2025/10
In parallel with the Advanced Technology Fuel (ATF) development program in Japan, Japan Atomic Energy Agency has established an R&D program on the ATF to provide more scientific support to the ATF fuel vendors in Japan. Based on the identified phenomenological issues, the R&D program covers the issues mainly on the light water reactors conditions such as normal operation, loss of coolant accident, beyond design basis accident, and severe accident. The R&D program such as irradiation test, corrosion test, LOCA test, and etc. are proposed. By acquiring the data from the experiment, the final goal is to implement the experimental data and model into the simulation code in order to predict the fuel performance behaviour at high burn-up. In 2024 TOPFUEL, the introduction of the R&D has been presented. In 2025, the update on the Cr-coated research to simulate normal operation and accident conditions will be given. For example, tests are currently underway to understand the in-reactor environment, such as ion irradiation, gamma irradiation and proton irradiation tests. In addition, a loss-of-coolant accident is being simulated for accident conditions, and the chemical effects of fission products on the cladding are being studied.
垣内 一雄; 成川 隆文*; 宇田川 豊; 勝山 仁哉; 三原 武; 天谷 政樹
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1440 - 1449, 2025/10
Phenomena of fuel fragmentation, relocation, and dispersal (FFRD) of high burnup light water reactor fuels have been observed under simulated loss of coolant accident (LOCA) experiments. If the fuel fragments accumulate densely in the ballooned cladding during LOCA, the power of fuel rod may increase locally, which may increase the peak cladding temperature. Furthermore, if a large number of fuel fragments were dispersed from fuel rod to reactor core, the coolability of reactor core during and after the accident may be influenced. While recent studies suggest large impact of rod internal gas state on fuel fragmentation and dispersal, there have been few experimental data that enable to evaluate such impact. We thus performed three LOCA-simulated burst tests (Test no. MMDA3 / MMDA4 / LZRT5) using irradiated PWR and BWR UO
fuel rods whose plenum volumes were designed to be 1 cc and 5 cc, respectively, as the main test parameter, at the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA). The tests highlighted the crucial role of plenum volume in fuel rod in FFRD: the burst appearance changed from a pin hole of MMDA3 with the 1cc plenum to a rupture opening of MMDA4 with 5 cc plenum, entailing increase in probable more substantial fragmentation and fuel fragments dispersal. Based on results from MMDA3 and MMDA4, the gas state, which was influenced by both the plenum volume and the gas communication, may significantly affect the amount of fuel fragment dispersion.
廣岡 瞬; Vauchy, R.; 堀井 雄太; 砂押 剛雄*; 齋藤 浩介
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 10 Pages, 2025/10
MOX燃料の酸素金属(O/M)比の低減は、燃料被覆管化学的相互作用(FCCI)による被覆管の腐食深さを抑制する上で重要な役割を果たし、高速炉MOX燃料の寿命を決定する鍵となる。MOXペレットの照射中、ペレット半径方向の温度勾配によって起こるO/M比の再分布のため、FCCIの発生箇所近傍であるペレット外周部では他の箇所より酸素ポテンシャルが高くなるものの、製造時のMOXペレットのO/M比を低く調整することで、腐食深さが明らかに減少する傾向が多くの照射試験および照射後試験により報告されている。MOXペレットの製造プロセスにおいてO/M比を低く調整するには、熱処理を高温、長時間に、そしてガス中の酸素分圧を低くすることで達成できると考えられるが、このO/M比の変化の特性は十分に研究されていない。本研究では、MOXペレットのO/M比を調整する試験を行い、熱分析装置を用いて熱処理中のO/M比の変化を評価した。これにより昇温中および高温保持中のO/M比の低下挙動と、冷却中のO/M比の上昇の挙動を明らかにした。さらに、FCCIに関して製造直後のO/M比よりも重要と思われるペレット外周部のO/M比を、Sariのモデルを用いて評価し、その結果からペレット外周部の酸素ポテンシャルを評価した。最新のMOXの酸素ポテンシャルのデータを用いることで、照射中のペレット半径方向の酸素ポテンシャルのプロファイルは外周部で特に急変化することが示され、また、この挙動は製造時のO/M比に敏感であることが明らかとなった。
谷口 良徳; 浦野 建太; 三原 武; 宇田川 豊; 垣内 一雄; 勝山 仁哉
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10
To investigate the fission gas release behavior of MOX fuel under reactivity-initiated accident (RIA) conditions, a RIA-simulated test on a high-burnup MOX fuel irradiated up to about 64.5 GWd/t (Test FGD-3) was conducted at the Nuclear Safety Research Reactor (NSRR) in JAEA by using recently developed Fission Gas Dynamics (FGD) testing technique. The concept of the FGD tests is to evaluate fission gas release during RIA-simulated test by measuring the pressure transient inside a rigid chamber containing the test fuel rod. We utilize Linear Variable Differential Transformer (LVDT)-type pressure sensor which less affected by gamma and/or neutron field in the NSRR core than conventional strain gauge-type pressure sensor. The maximum fuel enthalpy during Test FGD-3 was evaluated as 276 J/g, which is almost the same value as that of a previous FGD test on a high-burnup UO fuel (about 61 GWd/t) (Test FGD-2). The measured pressure increased from 0.1 MPa to eventually stabilized at about 0.75 MPa: this increase of pressure roughly corresponds to a transient FGR of about 28%, which is higher than that obtained in Test FGD-2 (about 18%). Sensitivity analyses of effective gas permeability for axial gas communication inside the FGD-3 test fuel rod using fuel performance code RANNS showed that apparent gas permeability of the FGD-3 fuel was much higher than that of the FGD-2 fuel. These results suggest that transient fission gas release from high-burnup MOX fuel exceeds that from UO fuel with similar burnup levels, and a significant portion released shortly after energy injection.
