Journal of Nuclear Materials最新文献_第4页

BISON benchmarking of the EBR-II X501 experiment for minor actinide-bearing metal fuel performance assessment BISON基准测试EBR-II X501实验用于微量锕系元素金属燃料性能评估

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-29 DOI: 10.1016/j.jnucmat.2025.156349

N.J. Fassino, E. Lacroix, W. Gillett, T. Arment, P. Everett

Metallic uranium-transuranic (U-TRU) fuels are promising candidates for advanced nuclear reactors, offering significant potential to reduce long-lived radiotoxic waste and close the nuclear fuel cycle. Qualification of these fuels for commercial deployment remains a challenge due to limited prototypic irradiation data and limited maturity of minor actinide material models. This study leverages detailed irradiation data and post-irradiation examination (PIE) results from the Experimental Breeder Reactor II (EBR-II) X501 experiment to develop and validate a high-fidelity fuel performance model using the BISON finite-element code, integrated with the Fuels Irradiation and Physics Database. To evaluate the influence of different minor-actinide approximations, three U-TRU modeling cases were implemented that respectively treated neptunium and americium as (A) plutonium, (B) uranium and zirconium, and (C) uranium. These cases enable assessment of how different thermophysical representations affect fuel performance predictions on an engineering scale. Key BISON input parameters affecting fuel performance underwent a multi-stage calibration. Initially, these parameters were adjusted to match PIE measurements from the minor actinide-bearing X501 pins, G582 and G591. As the analysis progressed and similarities between U-TRU and ternary fuels became clearer, the model was further refined using a broader set of historical PIE data from various ternary fuel pins irradiated in EBR-II. This comprehensive calibration strategy produced predictions in agreement with experimental observations, demonstrating robust predictive capability across varied alloy compositions and smeared densities. The findings support the argument that existing ternary fuel models in BISON can reasonably approximate the behavior of minor actinide-bearing fuels under fast reactor conditions. This study underscores the viability of utilizing limited but high-quality experimental data, such as X501 PIE, to accelerate qualification and licensing of minor actinide-bearing metallic fuels, supporting the deployment of sustainable nuclear fuel recycling technologies.

金属铀-超铀（U-TRU）燃料是先进核反应堆的有希望的候选燃料，具有减少长寿命放射性有毒废物和关闭核燃料循环的巨大潜力。由于原型辐照数据有限和次要锕系物质模型的成熟度有限，对这些燃料进行商业部署的资格鉴定仍然是一个挑战。本研究利用实验增殖反应堆II (EBR-II) X501实验的详细辐照数据和辐照后检查（PIE）结果，利用BISON有限元代码开发和验证高保真燃料性能模型，并集成燃料辐照和物理数据库。为了评估不同次锕系元素近似的影响，我们实施了三个U-TRU模型案例，分别将镎和镅处理为(A)钚，(B)铀和锆，以及(C)铀。这些案例能够在工程规模上评估不同的热物理表征如何影响燃料性能预测。影响燃料性能的关键BISON输入参数进行了多级校准。最初，对这些参数进行了调整，以匹配来自含锕系元素的X501引脚、G582和G591的PIE测量值。随着分析的深入，U-TRU和三元燃料之间的相似性变得更加清晰，使用EBR-II中辐照的各种三元燃料销的更广泛的历史PIE数据集进一步完善了该模型。这种全面的校准策略产生了与实验观察一致的预测，证明了对不同合金成分和涂抹密度的强大预测能力。这些发现支持了BISON现有的三元燃料模型可以合理地近似含少量锕系元素燃料在快堆条件下的行为的论点。这项研究强调了利用有限但高质量的实验数据（如X501 PIE）加速少量锕系元素金属燃料的鉴定和许可的可行性，支持可持续核燃料回收技术的部署。

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引用次数: 0

Effects of vacancies and voids on hydrogen behavior in yttrium hydrides 空位和空隙对氢化钇中氢行为的影响

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-29 DOI: 10.1016/j.jnucmat.2025.156350

Shuying Lin , Linbing Jiang , Baoliang Zhang , Xijun Wu , Siyi Shen , Yu Ma , Wenguan Liu

Yttrium hydride emerges as a promising high-temperature solid moderator for advanced nuclear reactors. Irradiation-induced vacancies and voids, which critically influence hydrogen (H) redistribution and moderation performance, remain not fully understood. This study systematically investigates the mechanistic role of Y vacancies on H migration through first-principles calculations and on-the-fly machine learning molecular dynamics simulations. First-principles analyses of binding energies and migration barriers reveal a pronounced repulsive interaction between Y vacancies and adjacent H atoms. This phenomenon arises from (i) weakened H binding at first nearest neighbor tetrahedral sites (T-sites) and enhanced binding at second nearest neighbor T-sites of the Y vacancy, and (ii) reduced migration barriers for H migration away from Y vacancies and elevated barriers for their reverse processes. Molecular dynamics simulations quantify H diffusion coefficients, radial distribution functions, H site occupancies, and spatial H distributions in systems containing Y vacancies/voids. Notably, H atoms are entirely excluded from the cores of Y vacancies/voids, even at elevated temperatures, precluding spontaneous formation of H bubbles or clusters in these irradiation-induced defects. Furthermore, Y vacancies/voids may exhibit resistance to macroscopic H transport under external driving forces, as evidenced by the Y vacancy/void-mediated H redistribution. These atomic-scale insights into interactions of Y vacancy and H aides to establish a theoretical framework for predicting H transport in irradiated yttrium hydrides moderators and elucidating the impact on neutron moderation efficiency.

氢化钇是一种很有前途的用于先进核反应堆的高温固体慢化剂。辐照引起的空位和空隙对氢(H)的再分配和调节性能有重要影响，但目前尚未完全了解。本研究通过第一性原理计算和动态机器学习分子动力学模拟系统地研究了Y空位在H迁移中的机制作用。结合能和迁移势垒的第一性原理分析揭示了Y空位和相邻H原子之间明显的排斥相互作用。这一现象源于(i) H在第一个最近四面体位点（t位点）的结合减弱，而在Y空位的第二个最近四面体位点的结合增强，以及（ii） H从Y空位迁移的迁移障碍降低，其反向过程的迁移障碍升高。分子动力学模拟量化了H扩散系数、径向分布函数、H位点占位率和含有Y空位/空隙的系统中的空间H分布。值得注意的是，即使在高温下，H原子也完全被排除在Y空位/空隙的核心之外，这就排除了在这些辐照诱导缺陷中自发形成H泡或团簇的可能性。此外，Y空位/空隙可能在外力作用下对宏观H输运表现出阻力，这可以通过Y空位/空隙介导的H重分布得到证明。这些对Y空位和H相互作用的原子尺度的见解有助于建立一个理论框架，用于预测辐照氢化钇减速剂中H的输运，并阐明对中子减速效率的影响。

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引用次数: 0

Modeling stress-driven diffusion of hydrogen in irradiated liner claddings using OFFBEAT 利用OFFBEAT模拟辐照衬层中氢的应力驱动扩散

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-29 DOI: 10.1016/j.jnucmat.2025.156351

P. Konarski , O. Yetik , L. Verma , I. Clifford , P. Trtik , L. Duarte , J. Bertsch , H. Ferroukhi

To ensure the safe long-term dry storage of used nuclear fuel in Switzerland, PSI investigates hydrogen behavior in Zircaloy liner claddings, commonly used in Swiss reactors. This study focuses on irradiated DX D4 cladding, consisting of a Zircaloy-4 substrate and an outer liner with low tin concentration, and examines hydrogen redistribution during thermo-mechanical treatments using the OFFBEAT fuel performance code. In this work, two irradiated samples of the cladding were simulated and compared to experimental results: one submitted to a thermal treatment, and the other additionally subjected to stress under mechanical loading. While the thermal-only case aligned with experimental data, the stress-influenced simulation diverged. To identify causes, parametric studies varying mechanical properties, hydrogen solubility (TSS), and diffusion parameters were carried out. Results indicate that TSS correlations have a greater impact than mechanical assumptions. Despite hydrogen profile mismatches in liner, predictions in the substrate remained robust. Moreover, the characteristic migration of hydrogen toward the liner is well captured. These findings contribute to understanding hydrogen behavior in duplex claddings and guide future model development for fuel performance under extended dry storage.

为了确保瑞士乏燃料的长期安全干储存，PSI调查了瑞士反应堆中常用的锆合金衬里包壳中的氢行为。本研究的重点是辐照的DX D4包层，由锆合金-4衬底和低锡浓度的外层组成，并使用OFFBEAT燃料性能代码检查热机械处理过程中的氢再分配。在这项工作中，对两个辐照后的包层样品进行了模拟，并与实验结果进行了比较：一个进行了热处理，另一个在机械载荷下进行了应力处理。当仅热情况与实验数据一致时，应力影响的模拟结果则不同。为了确定原因，进行了不同力学性能、氢溶解度（TSS）和扩散参数的参数研究。结果表明，TSS相关性比力学假设具有更大的影响。尽管在衬底中氢分布不匹配，但对衬底的预测仍然是稳健的。此外，氢气向衬里迁移的特征也被很好地捕捉到了。这些发现有助于理解氢在双包层中的行为，并指导未来在长时间干储存下燃料性能的模型开发。

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引用次数: 0

Epsilon phase simulants behavior during air oxidation and dissolution in nitric acid Epsilon相模拟了空气氧化和硝酸溶解过程中的行为

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-27 DOI: 10.1016/j.jnucmat.2025.156346

Mikhail Volgin , Andrey Shiryaev , Andrey Toropov , Sergey Kulyukhin , Konstantin Dvoeglazov , Iurii Nevolin

The epsilon phase (ε-phase) is an alloy comprising fission products such as Mo, Tc, Ru, Rh and Pd, abundant in insoluble residues of spent nuclear oxide fuel (SNF). In this study we address behavior of the simulant ε-phase and release of volatiles during high temperature voloxidation and dissolution in 8 M HNO₃. It is shown that oxidation in air starts at 650 °C; hexagonal lattice of the ε-phase is stable up to 1100 °C. Released volatile MoO₃ quantitatively precipitates already at 650 °C, while Tc and Ru oxides remain mobile down to room temperature. Dissolution of as-prepared samples was not completed after six days in 8 M HNO₃ at 95 °C. The presence of Tc in the simulated ε-phase strongly influences its microstructure, leaching kinetics and dissolution behavior. In particular, the Tc-containing simulated ε-phase possess rather uniform microstructure, whereas the Tc-free specimen is heterogeneous.

ε相（ε-相）是由Mo、Tc、Ru、Rh和Pd等裂变产物组成的合金，大量存在于乏氧核燃料（SNF）的不溶性残余物中。本文研究了模拟ε-相在8 M HNO3中高温氧化和溶解过程中挥发性物质的释放行为。结果表明，空气中的氧化始于650℃；ε-相的六角形晶格在1100℃时保持稳定。释放的挥发性MoO3在650℃时已经定量析出，而Tc和Ru氧化物在室温下仍保持流动。制备的样品在95°C下于8 M HNO3中溶解6天后未完全溶解。模拟ε-相中Tc的存在对其微观结构、浸出动力学和溶出行为影响较大。其中含tc的模拟ε相具有较为均匀的微观结构，而不含tc的模拟ε相具有不均匀的微观结构。

引用次数: 0

Tailoring aluminosilicates for optimized cesium removal through curing control strategy 定制铝硅酸盐通过固化控制策略优化铯去除

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-26 DOI: 10.1016/j.jnucmat.2025.156337

Vanessa Proust , Alban Gossard , Shivani Sharma , Scott T. Misture , Hans-Conrad zur Loye , Agnès Grandjean

Aluminosilicates are a class of materials interesting for the trapping of Cs⁺ from nuclear wastewater. The sorption properties of Cs⁺ by ionic exchange of such materials are widely depending on their structure and composition. From similar precursor’s solutions at different SiO₂/Al₂O₃ ratios, geopolymers and zeolitic structures (Na-P1 and Analcime, ANA) have been synthesized by ambient temperature curing and hydrothermal treatment respectively, and the Cs⁺ adsorption capacity of the materials has been evaluated. The Cs⁺ adsorption performances are directly controlled by the crystallinity and the microporosity of the materials, which influence the amount and the accessibility to mobile charge compensator cations. The porous structure of the NaP1 (predominantly obtained from SiO₂/Al₂O₃ ratios of 3 and 5) enables the adsorption of Cs⁺ by ionic exchange, while the channels of the analcime (synthesized from a SiO₂/Al₂O₃ ratio of 6) are too small for Cs⁺ to enter. Moreover, the presence of micropores, coming from defect in the crystalline structures or inter-crystallite spaces, facilitates the accessibility to the exchangeable sites in the materials and increases the sorption capacity. Thus, both the structure of the aluminosilicate solid phase and the overall porosity have to be considered to make accessible the largest possible amount of mobile charge compensator cations to improve at the best the material sorption capacity.

硅铝酸盐是一类对核废水中铯离子的捕获感兴趣的材料。这些材料的离子交换对Cs+的吸附性能在很大程度上取决于它们的结构和组成。以相似的前驱体溶液为原料，在不同SiO2/Al2O3配比下，通过常温固化和水热处理分别合成了具有Na-P1和Analcime （ANA）分子筛结构的地聚合物，并对材料的Cs+吸附能力进行了评价。Cs+吸附性能直接受材料的结晶度和微孔隙度的影响，从而影响移动电荷补偿阳离子的数量和可及性。NaP1（主要由SiO2/Al2O3比为3和5得到）的多孔结构使得Cs+可以通过离子交换吸附，而氧化铝（由SiO2/Al2O3比为6合成）的通道太小，Cs+无法进入。此外，微孔的存在，来自于晶体结构或晶间空间的缺陷，有利于材料中交换位的可达性，提高了吸附能力。因此，必须考虑铝硅酸盐固相的结构和整体孔隙率，以使尽可能多的移动电荷补偿器阳离子可用，从而在最佳情况下提高材料的吸附能力。

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引用次数: 0

The shear stress required for the splitting of a helical dislocation into a prismatic dislocation loop and a screw dislocation in alpha-iron α -铁中螺旋位错分裂为棱柱位错环和螺旋位错所需的剪切应力

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-26 DOI: 10.1016/j.jnucmat.2025.156339

Yosuke Tsunemoto , Tomohisa Kumagai , Kazuma Suzuki , Akiyoshi Nomoto , Atsuo Hirano , Akiyuki Takahashi

Reactor pressure vessel steels used in nuclear power plants become embrittled when exposed to neutrons emitted during nuclear fission. This is because of microstructural changes in the reactor pressure vessel steel, including the formation of prismatic dislocation loops. Previous studies have demonstrated that screw dislocations and prismatic dislocation loops interact to form helical dislocations. Therefore, this study investigates the resolved shear stress required for a helical dislocation to split into a prismatic dislocation loop and screw dislocation in alpha iron using molecular dynamics simulations. Simulations revealed that helical dislocations split into prismatic dislocation loops and screw dislocations under a threshold shear stress. The findings indicated that the threshold resolved shear stress decreased as the spacing between the dislocation loops increased and eventually converged. Furthermore, the threshold resolved shear stress increased with the radius of the helical dislocation loop. Two typical types of splitting behavior were observed: direct splitting with minimal bow-out and splitting via screw dislocation glide. A model was constructed to predict the threshold resolved shear stress as the maximum value between the stress required for direct splitting based on the energetics of the dislocation geometry and the stress based on the stress required for a screw dislocation to start glide. The model effectively described the trend of the threshold shear stress for helical dislocations.

核电站中使用的反应堆压力容器钢在暴露于核裂变过程中释放的中子中时会变脆。这是由于反应器压力容器钢的微结构变化，包括棱柱位错环的形成。先前的研究表明，螺旋位错和棱柱位错环相互作用形成螺旋位错。因此，本研究利用分子动力学模拟研究了α铁中螺旋位错分裂成棱柱位错环和螺旋位错所需的化解剪应力。模拟结果表明，在临界值剪应力作用下，螺旋位错可分为棱柱位错环和螺旋位错。结果表明，随着位错环间距的增大，阈值分解剪应力减小并最终收敛。随着螺旋位错环半径的增大，分解剪应力阈值也随之增大。观察到两种典型的劈裂行为：以最小弓形向外的直接劈裂和通过螺钉位错滑动的劈裂。建立了基于位错几何能量学的直接劈裂所需的应力与螺旋位错开始滑动所需的应力之间的最大值的阈值分解剪应力模型。该模型有效地描述了螺旋位错阈值剪应力的变化趋势。

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引用次数: 0

Impact of corrosion layer composition and thermal pretreatment on the radiolytic generation of molecular hydrogen from aluminum alloys 腐蚀层组成及热预处理对铝合金辐射分解生成分子氢的影响

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-26 DOI: 10.1016/j.jnucmat.2025.156336

Jacy K. Conrad , Corey D. Pilgrim , Hanna Hlushko , Elizabeth H. Parker-Quaife , Xiaofei Pu , Fidelma Giulia Di Lemma , Jagoda M. Urban-Klaehn , Fei Xu , Gregory P. Horne

Evaluating the technical feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled canisters requires accurately assessing the amount of radiation-induced molecular hydrogen gas (H₂) generated and the impact of absorbed radiation dose on the composition of ASNF corrosion layers. Here, we report a slowing in the rate of radiolytic H₂ generation from pre-corroded aluminum alloy (AA) 1100 and 6061 coupons irradiated up to 53 MGy of absorbed cobalt-60 gamma dose. By exploring a variety of thermal pretreatment conditions for the AA6061 coupons, we find that the “steady-state” yield of H₂ depends more on the aluminum corrosion layer composition than on the treatments applied prior to dry storage. Scanning electron microscopy and positron annihilation lifetime spectroscopy techniques provided evidence of defects in the corrosion layers of the investigated aluminum alloy coupons after gamma irradiation to high absorbed gamma doses (∼50 MGy), the consequences of which on fuel cladding integrity and H₂ generation should be explored in future works.

评价在氦气回填罐中延长干贮存铝包乏核燃料（ASNF）的技术可行性，需要准确评估辐射诱导分子氢气（H2）的生成量以及吸收的辐射剂量对ASNF腐蚀层组成的影响。在这里，我们报道了预腐蚀铝合金（AA） 1100和6061合金在吸收钴-60 γ剂量高达53 MGy的照射下产生放射性溶解H2的速率减慢。通过对AA6061合金的各种热处理条件的探索，我们发现H2的“稳态”产率更多地取决于铝腐蚀层的组成，而不是干燥储存前的处理。扫描电子显微镜和正电子湮灭寿命谱技术提供了高吸收伽马剂量（~ 50 MGy）照射后所研究的铝合金券腐蚀层存在缺陷的证据，其对燃料包层完整性和H2生成的影响应在未来的工作中进行探讨。

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引用次数: 0

Growth behavior of dislocation loops in FeCoNiCrCu high-entropy alloy FeCoNiCrCu高熵合金中位错环的生长行为

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-26 DOI: 10.1016/j.jnucmat.2025.156338

La Han, Chuanlong Xu, Peidong Li, Xiaobao Tian, Wentao Jiang, Qingyuan Wang, Haidong Fan

High-entropy alloys (HEAs) with superior irradiation resistance are highly promising for nuclear reactor applications. In this work, the growth behavior of prismatic dislocation loops and Frank dislocation loops in FeCoNiCrCu HEA and pure Ni was studied by molecular dynamics simulations. The simulation results showed that the prismatic loops and Frank loops in Ni and FeCoNiCrCu HEA are able to grow by absorbing IAs (interstitial atoms). The prismatic loops grow at the same rate as Frank loops in Ni and HEA at low temperature, but faster at room/high temperature. In addition, the dislocation loops in Ni always grow faster than those in HEA at all temperatures. The growth length of dislocation loops in Ni and HEA increases with the increasing IA concentration and loop radius. Regarding the growth mechanism, the IA absorption area around prismatic loops is larger than that around Frank loops. The IA absorption areas in pure Ni are always larger than those in HEA at all temperatures. The results in this work clearly show that the HEA exhibits higher resistance to loop growth than Ni.

高熵合金（HEAs）具有优异的抗辐照性能，在核反应堆领域具有广阔的应用前景。本文通过分子动力学模拟研究了棱柱位错环和Frank位错环在FeCoNiCrCu HEA和纯Ni中的生长行为。模拟结果表明，Ni和FeCoNiCrCu HEA中的棱柱形环和Frank环能够通过吸收间隙原子生长。在Ni和HEA中，棱柱形环在低温下的生长速率与Frank环相同，但在室温/高温下更快。此外，在所有温度下，Ni中的位错环都比HEA中的位错环生长得快。Ni和HEA中位错环的生长长度随着IA浓度和环半径的增加而增加。在生长机理上，棱柱形环周围的IA吸收面积大于Frank环周围的IA吸收面积。在任何温度下，纯Ni的IA吸收面积都大于HEA的吸收面积。结果清楚地表明，HEA比Ni具有更高的耐环生长能力。

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引用次数: 0

Post-irradiation examination of AGR-3/4 TRISO fuel compacts using three-dimensional X-ray computed tomography AGR-3/4 TRISO燃料包辐照后的三维x射线计算机断层扫描研究

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-26 DOI: 10.1016/j.jnucmat.2025.156341

Swapnil Morankar , William C. Chuirazzi , Rahul R. Kancharla , Josh J. Kane , Brian Gross , Nikolaus L. Cordes , John D. Stempien

The AGR-3/4 irradiation tests combined the third and fourth planned irradiation experiments in the US Department of Energy’s Advanced Gas Reactor (AGR) testing campaign of tri-structural isotropic (TRISO) fuel compacts. In this article, we present post-irradiation examination (PIE) using X-ray computed tomography (XCT) of two unirradiated and two irradiated compacts from the AGR-3/4 irradiation tests. The irradiated compacts studied (compact 7–1 and compact 12–4) represent the upper and lower limit of burnup within the AGR-3/4 irradiation experiment. This article presents a detailed quantitative analysis on the post-irradiation structure of TRISO fuel compacts. Various quantitative parameters including shape, size, and packing of kernels, and their spatial distribution, were utilized to gain insights into the structural changes caused by irradiation. The equivalent diameter and sphericity were found to increase and decrease, respectively, in irradiated compact 7–1 due to its higher burnup. Nearest neighbor distance between fuel kernels decreased after irradiation, suggesting irradiation-induced shrinkage of graphitic matrix. Furthermore, each compact in AGR-3/4 irradiation tests contained 20 designed-to-fail (DTF) fuel particles that were meant to act as a source of fission product release to the experiment test train. In the present work, all DTF fuel particles in the four compacts studied were identified, and it was found that they exhibited larger kernel swelling in compact 12–4 and smaller kernel swelling in compact 7–1, compared to the driver particles.

AGR-3/4辐照试验结合了美国能源部先进气体反应堆（AGR）三结构各向同性（TRISO）燃料紧凑型试验计划的第三次和第四次辐照试验。在本文中，我们使用x射线计算机断层扫描（XCT）对两个未辐照和两个辐照后的AGR-3/4辐照试验的压缩物进行辐照后检查（PIE）。所研究的辐照紧凑体（紧凑7-1和紧凑12-4）代表AGR-3/4辐照实验的燃耗上限和下限。本文对TRISO燃料棒的辐照后结构进行了详细的定量分析。利用各种定量参数，包括核的形状、大小和包装及其空间分布，来深入了解辐照引起的结构变化。结果表明，7-1型辐照后的等效直径增大，球度减小，燃耗增大。辐照后燃料核之间的最近邻距离减小，表明辐照引起石墨基体收缩。此外，AGR-3/4辐照试验中的每个紧凑型包含20个设计失效（DTF）燃料颗粒，旨在作为裂变产物释放到实验测试列车的来源。在本工作中，我们对四种压缩颗粒中的DTF燃料颗粒进行了识别，发现与驱动颗粒相比，压缩颗粒12-4的核膨胀较大，压缩颗粒7-1的核膨胀较小。

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引用次数: 0

Coupled EBSD/TKD/APT study of segregation induced by irradiation in a Fe-13at.%Cr model alloy through different grain boundaries type Fe-13at辐照诱导偏析的耦合EBSD/TKD/APT研究。%Cr模型合金通过不同的晶界类型

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-11-25 DOI: 10.1016/j.jnucmat.2025.156334

Q. Barrès , O. Tissot , E. Meslin , I. Mouton , M. Loyer-Prost , C. Pareige

The 4th generation of nuclear reactors currently under study is based on fast neutrons. The high energy of neutrons induces new constraints for the surrounding structural materials such as the reactor vessel and the cladding. Materials capable of withstanding the new operating conditions must be found. Ferritic/Martensitic and ODS steels are good candidates for addressing corrosion, mechanical and irradiations issues. Under irradiation, the creation and migration of point defects leads to various mechanisms that modify the initial properties of materials. Radiation induced segregation (RIS) is one of these mechanisms. RIS will occur based on various parameters related to materials and the radiative environment like temperature or dose. This paper presents the quantification of RIS on different types of grain boundaries (GB) in an FeCr model alloy. Correlative analyses before and after irradiation have been conducted on the same GB structure using Electron Backscatter Diffraction (EBSD) and Transmission Kikuchi Diffraction (TKD) techniques. Chemical quantifications were performed using Atom Probe Tomography (APT). Systematic W-shape Cr profile across GB after irradiation is revealed. The extent of this phenomenon depends on the structure of the GB being studied.

目前正在研究的第四代核反应堆是基于快中子的。中子的高能量对反应堆容器和包层等周围结构材料产生了新的约束。必须找到能够承受新的操作条件的材料。铁素体/马氏体钢和ODS钢是解决腐蚀、机械和辐照问题的良好候选者。在辐照下，点缺陷的产生和迁移导致了各种改变材料初始性能的机制。辐射诱导偏析（RIS）是其中一种机制。RIS将根据与材料和辐射环境（如温度或剂量）相关的各种参数发生。本文对一种FeCr模型合金中不同类型晶界的RIS进行了定量分析。利用电子背散射衍射（EBSD）和透射菊池衍射（TKD）技术对同一GB结构辐照前后进行了相关分析。使用原子探针断层扫描（APT）进行化学定量。辐照后在GB上显示出系统的w形Cr分布。这种现象的程度取决于所研究的GB的结构。

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引用次数: 0