Sintering behavior analysis of compacted dry recycled UPuO powder using master sintering curve theory

中道 晋哉; 砂押 剛雄*; 廣岡 瞬; Vauchy, R.; 村上 龍敏

Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07

https://doi.org/10.1016/j.jnucmat.2024.155072

Using dry recycled powders for uranium and plutonium mixed oxide (MOX) fuel production can reduce unnecessary storage and accountability of nuclear material in facilities. The shrinkage behavior of green compacts of dry recycled powders differs from that of conventional raw powders because the dry recycled MOX powder is obtained from the fabrication scrap of sintered pellets. The shrinkage behavior of dry recycled MOX powder has been investigated by dilatometry. Based on the shrinkage curves, sintering apparent activation energies were evaluated using the master sintering curve (MSC) and the constant rate of heating methods. The obtained values were higher than the energy evaluated for raw powder experiments. The sigmoid sintering prediction equation using the MSC function was constructed. The accumulation of data on the activation energy for various sintering conditions will lead to the wide application of this prediction formula in the future.

Oxidation and embrittlement behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之

Journal of Nuclear Materials, 587, p.154736_1 - 154736_8, 2023/12

https://doi.org/10.1016/j.jnucmat.2023.154736

To evaluate the oxidation and embrittlement behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions, we conducted isothermal oxidation and ring-compression tests on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens. Further, we discussed the loss of coolable geometry of the reactor core loaded with the FeCrAl-ODS cladding tubes under LOCA conditions, using data from the ring-compression tests in this study and the integral thermal shock tests from our previous study. The results reveal that oxidation kinetics of the FeCrAl-ODS cladding tube at 1523 K is four orders of magnitude lower than that of a conventional Zircaloy cladding tube, which highlights the exceptional oxidation resistance of the FeCrAl-ODS cladding tube. The breakaway oxidation of the FeCrAl-ODS cladding tube was observed at 1623 K for durations equal to or exceeding 6 h, and melting was observed at 1723 K. The ring-compression and the integral thermal shock tests indicate that, depending on the oxidation time, the ductile to brittle transition threshold - as determined by the ring-compression test - exists between 1623 K and 1723 K. Meanwhile, the fracture threshold - established through the integral thermal shock test - falls between 1573 K and 1673 K. Therefore, taking a conservative approach based on available data, the fracture and non-fracture results from the integral thermal shock tests can define the lower and upper boundaries of the threshold for the loss of coolable geometry of the reactor core during a LOCA.

中性子回折によるハイエントロピー合金の変形挙動のその場観察

Harjo, S.

日本結晶学会誌, 65(3), p.178 - 182, 2023/08

https://doi.org/10.5940/jcrsj.65.178

Observations of deformation behavior of high entropy alloys using neutron diffraction measurements during deformation at various temperatures are reviewed. Neutrons are used to investigate stresses and crystallographic microstructures inside engineering materials, taking advantage of their large penetrating power and the ability to see the arrangement of atoms by diffraction methods. The important structural details of high entropy alloys such as internal stresses, phase conditions, dislocations, texture etc. are discussed in relation to the deformation conditions. Some highlights are introduced: (a) Cooperative deformation in CrMnFeCoNi alloy at ultralow temperatures, (b) Stacking fault energies in CrFeCoNi and CrCoNi alloys, and (c) Load redistribution in eutectic high entropy alloy AlCoCrFeNi during high temperature deformation.

Large-eddy simulation on two-liquid mixing in the horizontal leg and downcomer (the TAMU-CFD Benchmark), with respect to fluctuation behavior of liquid concentration

安部 諭; 岡垣 百合亜

Nuclear Engineering and Design, 404, p.112165_1 - 112165_14, 2023/04

https://doi.org/10.1016/j.nucengdes.2023.112165

Pressurized Thermal Shock (PTS) is induced potentially by the rapid cooling of the cold-leg and downcomer wall in the primary system of a Pressurized Water Reactor (PWR) due to the initiation of Emergency Core Cooling System (ECCS). Thus, fluids mixing in a horizontal cold-leg and downcomer should be predicted accurately; however, turbulence production and damping often hinders this prediction due to the presence of the density gradients. Hence, the Fifth International Benchmark Exercise, the cold-leg mixing Computational Fluid Dynamics (CFD) Benchmark, was conducted under the support of OECD/NEA. The experiment was designed for visualization of the mixing phenomena of two liquids with different densities. The heavy liquid was a simulant of cold water from ECCS, in a horizontal leg and downcomer. We used the Large-eddy Simulation (LES) to investigate the time fluctuation behaviors of velocity and liquid concentration. The CFD simulation was performed with two turbulence models and three different numerical meshes. We investigated the characteristics of the appearance frequency of the heavy liquid concentration with the statistical method. Based on our findings, we propose further experiments and numerical investigations to understand the fluid mixing phenomena related to PTS.

Particle-based simulation of jet impingement behaviors

高塚 大地*; 守田 幸路*; Liu, W.*; Zhang, T.*; 中村 武志*; 神山 健司

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 10 Pages, 2022/10

A 3D particle-based simulation code was developed to analyze jet impingement behavior, and the physical models for thermal-hydraulic interactions between molten jets and solid plates used in the code was validated by simulating existing jet impingement experiments. In addition, particle-based simulations were conducted to understand the impingement characteristics of molten MOX fuel jets on SS plates, and the erosion rate of the SS plate was evaluated. The results showed that fuel crusts formed on the plate potentially provide thermal protection and confirmed the effect of the plate erosion rate on the jet diameter under conditions where crust formation is dominant.

Quantum critical spin-liquid-like behavior in the = quasikagome-lattice compound CeRhPdSn investigated using muon spin relaxation and neutron scattering

Tripathi, R.*; Adroja, D. T.*; Ritter, C.*; Sharma, S.*; Yang, C.*; Hillier, A. D.*; Koza, M. M.*; Demmel, F.*; Sundaresan, A.*; Langridge, S.*; et al.

Physical Review B, 106(6), p.064436_1 - 064436_17, 2022/08

https://doi.org/10.1103/PhysRevB.106.064436

We present the results of muon spin relaxation (SR) and neutron scattering measurements on the Ce-based quasikagome lattice CeRhPdSn ( = 0.1 to 0.75). Our SR results reveal the absence of both static long-range magnetic order and spin freezing down to 0.05 K in the = 0.1 sample, which is consistent with neutron scattering results. The weak temperature-dependent plateaus of the dynamic spin fluctuations below 0.2 K in zero field together with its longitudinal-field (LF) dependence indicate the presence of dynamic spin fluctuations persisting even at 0.05 K without static magnetic order. We argue that such a behavior for = 0.1 can be attributed to a metallic spin-liquid-like ground state near the quantum critical point in the frustrated Kondo lattice. The LF-SR study suggests that the out of kagome plane spin fluctuations are responsible for the spin-liquid behavior.

燃料デブリ取出し時における放射性核種飛散防止技術の開発(委託研究); 令和2年度英知を結集した原子力科学技術・人材育成推進事業

廃炉環境国際共同研究センター; 東京大学*

JAEA-Review 2022-010, 155 Pages, 2022/06

JAEA-Review-2022-010.pdf:9.78MB

日本原子力研究開発機構(JAEA)廃炉環境国際共同研究センター(CLADS)では、令和2年度英知を結集した原子力科学技術・人材育成推進事業(以下、「本事業」という)を実施している。本事業は、東京電力ホールディングス株式会社福島第一原子力発電所の廃炉等をはじめとした原子力分野の課題解決に貢献するため、国内外の英知を結集し、様々な分野の知見や経験を、従前の機関や分野の壁を越えて緊密に融合・連携させた基礎的・基盤的研究及び人材育成を推進することを目的としている。平成30年度の新規採択課題から実施主体を文部科学省からJAEAに移行することで、JAEAとアカデミアとの連携を強化し、廃炉に資する中長期的な研究開発・人材育成をより安定的かつ継続的に実施する体制を構築した。本研究は、研究課題のうち、平成30年度に採択された「燃料デブリ取出し時における放射性核種飛散防止技術の開発」の平成30年度から令和3年度の研究成果について取りまとめたものである(令和3年度まで契約延長)。本課題は令和3年度が最終年度となるため4年度分の成果を取りまとめた。本研究は、東京電力ホールディングス株式会社福島第一原子力発電所の燃料デブリ取出し時に発生する放射性微粒子を着実に閉じ込めるために、気相及び液相における微粒子の挙動を明らかにするとともに、微粒子飛散防止技術を開発することを目的としている。具体的には、微粒子飛散防止技術として、(1)スプレー水等を利用して最小限の水で微粒子飛散を抑制する方法、及び(2)燃料デブリを固化して取り出すことで微粒子飛散を抑制する方法について、実験及びシミュレーションにより評価し、実機適用可能性を評価した。

Development of PHITS plugin for calculating the integral radiation dose on robots

鈴木 健太; 川端 邦明

New Trends in Intelligent Software Methodologies, Tools and Techniques; Frontiers in Artificial Intelligence and Applications, Vol.355, p.248 - 257, 2022/00

https://doi.org/10.3233/FAIA220255

This paper describes the development of PHITS Plugin for calculating an integral radiation dose. PHITS Plugin is an extension of Choreonoid, and it was developed to estimate the effects of radiation on robots by calculating the integral radiation dose. We discuss the functional requirements to compute the dose distribution for calculating the integral radiation dose during the simulation.

Development of an integrated computer code system for analyzing irradiation behaviors of a fast reactor fuel

上羽 智之; 根本 潤一*; 伊藤 昌弘*; 石谷 行生*; 堂田 哲広; 田中 正暁; 大塚 智史

Nuclear Technology, 207(8), p.1280 - 1289, 2021/08

https://doi.org/10.1080/00295450.2020.1810977

高速炉燃料集合体の冷却材熱流動、燃料ピンの照射挙動、燃料ピン束の照射変形を連成して解析する統合計算コードシステムを開発した。このシステムは複数の計算コードから構成され、各コードが計算に必要とする情報を他のコードの計算結果から得るようになっている。これにより、照射下の燃料集合体における熱,機械,化学的挙動を関連させて解析することができる。本システムの機能確認のテスト解析として、高速炉で照射した混合酸化物燃料ピン束集合体の照射挙動解析を実施した。解析結果は集合体の横断面図、集合体や燃料ピンの3次元イメージモデル上に描画した。更に、解析で得られた燃料ピンの様々な照射挙動について、照射条件の影響を評価した。

Leaching behavior of radionuclides from samples prepared from spent fuel rod comparable to core debris in the 1F NPS

大西 貴士; 前田 宏治; 勝山 幸三

Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04

https://doi.org/10.1080/00223131.2020.1822943

To investigate the leaching behavior of radioactive nuclides in leaching samples comparable to core debris (partially molten ZrO/UO between fuel rods) in 1F NPS, the concentration of radionuclides in the leaching solution was measured. Leaching behaviors of actinides (U, Pu, Np) and Cs from the samples were similar to those from spent fuel. Leaching of U and Pu depends on pH in the cooling water of the core debris as predicted from the present thermodynamic database. While, if Mo and Tc are surrounded by zircaloy in the core debris, their leaching amount may become higher by one order of magnitude than those from spent fuel.