Nuclear Materials and Energy最新文献_第6页

Direct experimental measurement of the solubility of hydrogen isotopes in the Eurofer-97 RAFM 氢同位素在Eurofer-97 RAFM中溶解度的直接实验测量

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-03-01 Epub Date: 2026-02-23 DOI: 10.1016/j.nme.2026.102090

Igor Peñalva , María Urrestizala , Natalia Alegría , Marcelo Roldán , Marta Malo

Accurate knowledge of hydrogen isotopes transport properties in structural materials of interest for Breeding Blankets (BB) and Tritium Extraction and Recovery (TER) systems is fundamental for experiments, modelling, safety and design activities. This work aims at measuring these transport properties in Eurofer-97 at laboratory conditions, providing a reliable measure, at relevant operational conditions, on which a degree of uncertainty exists.

The solubility of protium and deuterium, expressed in terms of the Sieverts’ constant (K_S), is commonly derived indirectly from permeation experiments, assuming Sieverts’ law and diffusion-limited regimen (DLR). In this work, aiming the goal of avoiding these indirect procedure, solubility of protium and deuterium in Eurofer-97 was directly measured by means of the gas evolution absorption–desorption technique at the facility located in the University of the Basque Country (EHU). This facility has been previously used and validated for the direct determination of hydrogen isotope solubility in other reduced-activation ferritic–martensitic steels (RAFM).

Absorption–desorption experiments were carried out using Eurofer-97 samples at three different temperatures, for both protium and deuterium. The resulting values of Sieverts’ constants expressed in [mol m⁻³ Pa^−0.5] for protium at have been 8.75 × 10⁻³ at 350 °C, 1.13 × 10⁻² at 450 °C and 1.33 × 10⁻² at 550 °C, while for deuterium values of 7.88 × 10⁻³, 1.15 × 10⁻², and 1.26 × 10⁻² have been obtained at the same respective temperatures. Within the experimental uncertainty, no clear isotopic effect on solubility can be confirmed in the temperature range investigated, which is consistent with classical transport theory. Likewise, the measured values are in good agreement with literature data, especially with the most recent results. This study provides, for the first time for Eurofer-97, a fully direct experimental determination of the solubility of hydrogen isotopes, which contributes to reducing uncertainties in the modelling of hydrogen transport for fusion applications.

在繁殖毯（BB）和氚提取和回收（TER）系统中，对结构材料中氢同位素输运特性的准确了解是实验、建模、安全和设计活动的基础。这项工作的目的是在实验室条件下测量Eurofer-97的这些输运特性，在存在一定程度不确定性的相关操作条件下提供可靠的测量。用Sieverts常数（KS）表示的protium和deuterium的溶解度，通常是从渗透实验中间接得出的，假设Sieverts定律和扩散限制方案（DLR）。在这项工作中，为了避免这些间接过程，在位于巴斯克大学（EHU）的设施中，通过气体释放吸收-解吸技术直接测量了protium和氘在Eurofer-97中的溶解度。该设备先前已用于直接测定氢同位素在其他低活化铁素体-马氏体钢（RAFM）中的溶解度。利用Eurofer-97样品在三种不同的温度下对protium和deuterium进行了吸附-解吸实验。在350℃时，质子的Sieverts常数为8.75 × 10−3，在450℃时为1.13 × 10−2，在550℃时为1.33 × 10−2，而在相同的温度下，氘的Sieverts常数为7.88 × 10−3,1.15 × 10−2，和1.26 × 10−2。在实验的不确定度范围内，在所研究的温度范围内，没有明确的同位素对溶解度的影响，这与经典输运理论一致。同样，测量值与文献数据，特别是与最近的结果非常吻合。这项研究首次为Eurofer-97提供了氢同位素溶解度的完全直接实验测定，这有助于减少聚变应用中氢输运建模的不确定性。

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引用次数: 0

Effect of the angle of incidence of He, T, and D ions irradiation on physical and chemical sputtering of graphite targets in the near sputtering threshold energy regime He、T和D离子辐照入射角对近溅射阈能下石墨靶物理和化学溅射的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2026-03-01 Epub Date: 2025-12-12 DOI: 10.1016/j.nme.2025.102047

Al-Montaser Bellah A. Al-Ajlony , Ghadeer H. Al-Malkawi

This study presents a comprehensive Monte Carlo simulation-based investigation into the angle of incidence dependence of physical and chemical sputtering of graphite targets irradiated by He⁺, T⁺, and D⁺ ions in the near-threshold energy regime. This investigation has been executed by utilizing two advanced simulation codes, RDS-BASIC and SDTrimSP. In this study, we assess sputtering yields and energy thresholds across incidence angles ranging from 0° to 80°. Results indicate that physical sputtering yields are maximized at grazing angles (60°–80°), accompanied by a systematic decrease in sputtering threshold energy with increasing angle. In contrast, chemical sputtering, modeled for T⁺ and D⁺ ions, reveals a two-step threshold behavior: an initial erosion onset at ∼ 5 eV, and a secondary enhancement between 8–13 eV driven by physical displacement effects. Which also was found to influence the angle og incidence dependence of the chemical sputtering. Thereby, to physical sputtering, chemical erosion exhibits a peak at intermediate angles (60°–70°). These findings offer key insights into the erosion mechanisms of plasma-facing components and support the optimized design of carbon-based materials for future nuclear fusion reactor applications.

本研究基于蒙特卡罗模拟全面研究了He+， T+和D+离子在近阈值能量状态下辐照石墨靶的物理和化学溅射的入射角依赖关系。本研究利用两个先进的仿真代码RDS-BASIC和SDTrimSP进行。在这项研究中，我们评估了从0°到80°的入射角范围内的溅射产量和能量阈值。结果表明：在60°~ 80°的掠射角范围内，物理溅射产量最大，溅射阈值能量随掠射角的增大而降低；相反，化学溅射，模拟T+和D+离子，揭示了两步阈值行为：初始侵蚀开始于~ 5 eV，在8-13 eV之间由物理位移效应驱动的二次增强。这也影响了化学溅射的入射角依赖性。因此，对于物理溅射，化学侵蚀在中间角度（60°-70°）出现峰值。这些发现为了解等离子体组件的侵蚀机制提供了关键见解，并为未来核聚变反应堆应用的碳基材料的优化设计提供了支持。

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引用次数: 0

Role of doped ZrC on deuterium trapping in W-ZrC alloy 掺杂ZrC对W-ZrC合金中氘俘获的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-11-10 DOI: 10.1016/j.nme.2025.102023

Xuexi Zhang , Li Qiao , Hong Zhang , Xuefeng Xie , Yange Zhang , Peng Wang , Changsong Liu

Understanding and predicting hydrogen isotopes (His) retention in plasma-facing materials (PFMs) is crucial for the safe and efficient operation of fusion reactors. Here, a new candidate PFM, W-ZrC alloy, was exposed to D plasma at temperatures in the range from 400 K to 850 K. Surface morphology analysis revealed that the maximum blistering effect on W-ZrC alloy occurred at 600 K, which is 100 K higher than for pure tungsten (W). A quantitative statistical analysis revealed that higher temperature led to an increase in blister diameter and a concomitant decrease in areal density for both W and W-ZrC alloy. The blisters on the W and W-ZrC alloy originated from subsurface cavities, with nucleation sites localized in both intragranular and intergranular regions. The intergranular cavity beneath the blisters of W-ZrC alloy tended to extend along the phase boundaries between ZrC particles and W grains, and the ZrC particles at grain boundaries effectively suppressed the intragranular cavity propagation. If the exposure temperature exceeded 500 K, implanted D atoms dissolved and accumulated within ZrC grains in the W-ZrC alloy, leading to a new high-temperature thermal desorption spectroscopy (TDS) peak ∼ 980 K. Additionally, the total D retention in W-ZrC alloy is higher than that in pure W, especially at the exposure temperatures of 600 K and 700 K. This work provides key insights into surface blistering and D retention behavior in W-ZrC alloys, establishing a foundational basis for optimizing their performance as PFMs in fusion reactor applications.

了解和预测等离子体材料（PFMs）中氢同位素（His）的保留对核聚变反应堆的安全高效运行至关重要。在这里，一种新的候选PFM， W-ZrC合金，在400 K到850 K的温度范围内暴露在D等离子体中。表面形貌分析表明，W- zrc合金在600 K时起泡效果最大，比纯钨(W)高100 K。定量统计分析表明，温度升高导致W和W- zrc合金的泡口直径增大，面密度减小。W和W- zrc合金上的水泡起源于亚表面空腔，晶内和晶间均有形核。W-ZrC合金水泡下的晶间空洞倾向于沿ZrC颗粒与W晶粒的相界扩展，晶界处的ZrC颗粒有效地抑制了晶内空洞的扩展。当暴露温度超过500 K时，注入的D原子溶解并积聚在W-ZrC合金的ZrC晶粒内，形成一个新的高温热脱附光谱（TDS）峰~ 980 K。此外，W- zrc合金中总D的保留量高于纯W，特别是在600 K和700 K的暴露温度下。这项工作为W-ZrC合金的表面起泡和D保留行为提供了关键见解，为优化其在聚变反应堆应用中的pfm性能奠定了基础。

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引用次数: 0

Influence of redeposited tungsten and EUROFER97 layers on deuterium retention in plasma-facing materials 再沉积钨和EUROFER97层对等离子体材料中氘潴留的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-09-20 DOI: 10.1016/j.nme.2025.101990

Martina Fellinger , Eduardo Pitthan , Daniel Gautam , Daniel Primetzhofer , Friedrich Aumayr

Retention of hydrogen isotopes in plasma-facing materials is a key challenge for safety and fuel efficiency of nuclear fusion reactors. In realistic reactor environments, simultaneous processes, such as erosion, redeposition, implantation and outgassing, can alter surface compositions and may affect hydrogen isotope retention. In this study, we investigate how thin redeposited layers of tungsten and EUROFER97 influence retention and release of previously implanted deuterium. Using a combination of Elastic Recoil Detection Analysis and Rutherford Backscattering Spectrometry, we quantify deuterium retention during in-situ annealing up to 600 °C. Comparisons between coated and uncoated samples show that redeposited tungsten can act as partial diffusion barrier, preventing deuterium from outgassing. In contrast, redeposited EUROFER97 layers show no such effect and appear virtually transparent to deuterium diffusion. These findings emphasize the critical role of redeposited layers on fuel retention and have implications for wall lifetime estimates and fuel inventory control in fusion devices.

氢同位素在等离子体材料中的保留是核聚变反应堆安全性和燃料效率的关键挑战。在现实的反应堆环境中，同时发生的过程，如侵蚀、再沉积、注入和放气，可以改变表面成分，并可能影响氢同位素的保留。在这项研究中，我们研究了再沉积的薄层钨和EUROFER97如何影响先前植入的氘的保留和释放。使用弹性反冲检测分析和卢瑟福后向散射光谱相结合，我们量化了在高达600°C的原位退火过程中的氘保留量。对包覆和未包覆样品的比较表明，再沉积的钨可以起到部分扩散屏障的作用，阻止氘的脱气。相比之下，重新沉积的EUROFER97层没有这种效应，对氘扩散几乎是透明的。这些发现强调了再沉积层对燃料保留的关键作用，并对核聚变装置的壁寿命估计和燃料库存控制具有重要意义。

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引用次数: 0

Understanding the oxidation of pure tungsten in air and its impact on the lifecycle of a fusion power plant 了解纯钨在空气中的氧化及其对核聚变电厂生命周期的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-09-18 DOI: 10.1016/j.nme.2025.101988

Rongrui Li , Guillermo Álvarez , Ayla Ipakchi , Livia Cupertino-Malheiros , Mark R. Gilbert , Emilio Martínez-Pañeda , Eric Prestat

The oxidation of pure W and the sublimation of W oxide have been investigated to assess their impact on the lifecycle of a fusion power plant. Pure W has been oxidised at temperatures between 400 and 1050 °C and for durations ranging between 1 and 70 h. The formation of voids and cracks has been observed at temperatures above 600 °C, leading to the formation of dust or oxide spalling, which could be problematic in maintenance and waste-handling scenarios of a fusion power plant. Preferential oxidation taking place at the edge of the specimen was characterised, and its impact is discussed in relation to component design. Characterisation using electron microscopy and Raman spectroscopy revealed that the oxide scale is formed of three main layers: the inner layer is 30–50 nm thick WO

_{2}

oxide, the middle layer is a 10–20

μ

m thick of WO_2.72 and the outer layer is formed of WO_2.9/WO

_{3}

phases — whose thickness varies according to the total thickness of the oxide scale. The observed microstructure is discussed in relation to the parabolic-to-linear kinetics and its potential impact on tritium permeation and detritiation efficiency.

研究了纯W的氧化和W氧化物的升华，以评估它们对核聚变电厂生命周期的影响。纯W在400至1050°C的温度下氧化，持续时间在1至70小时之间。在600°C以上的温度下观察到空洞和裂纹的形成，导致灰尘或氧化物剥落的形成，这可能会在核聚变发电厂的维护和废物处理场景中产生问题。优先氧化发生在试样的边缘进行了表征，并讨论了其影响有关的组件设计。利用电子显微镜和拉曼光谱对氧化层进行了表征，发现氧化层主要由三层组成：内层为30 ~ 50 nm厚的WO2氧化物，中间层为10 ~ 20 μm厚的WO2.72，外层为WO2.9/WO3相，其厚度根据氧化层的总厚度而变化。讨论了观察到的微观结构与抛物线-线性动力学的关系及其对氚渗透和除氚效率的潜在影响。

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引用次数: 0

Improving accuracy in fluoride salt composition analysis: A focus on sample preparation for ICP-MS 提高氟化物组成分析的准确性：ICP-MS样品制备的重点

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 DOI: 10.1016/j.nme.2025.102039

Nayoung Kim , Weiyue Zhou , Kevin B. Woller , Alexander A. Khechfe , Guiqiu Zheng , Caroline Sorensen , Michael P. Short

Molten salts are versatile candidates for heat transfer/storage due to attractive thermophysical, thermochemical, and radiochemical properties. Salt impurities are a key factor in these properties, affecting corrosion of salt-facing materials and activation in nuclear applications. Inductively coupled plasma mass spectrometry (ICP-MS) can be used to measure metallic impurities in salts to high precision. However, there is no universally accepted method for ICP-MS analysis of salt impurities comparable to well-established methodologies for metals. In this paper, we present how ICP-MS analysis of (LiF)_0.465-(NaF)_0.115-(KF)_0.42 (FLiNaK) depends heavily on each step of sample preparation. Different sampling methods, salt particle sizes, dissolving acid concentrations, digestion methods, and microwave digestion conditions are explored. The optimal experimental condition for each is discussed with insights on safety, time management, and unit choice for seamless communication, and verified by measuring intentionally added metallic impurities (MnF₂, NiF₂, CoF₂, FeF₂). This study presents a repeatable method for conducting accurate ICP-MS measurements on salt to yield more consistent and comparable data.

熔融盐由于其吸引人的热物理、热化学和放射化学性质，是热传递/存储的多功能候选材料。盐杂质是影响这些性能的一个关键因素，它会影响临盐材料的腐蚀和核应用中的活化。电感耦合等离子体质谱法（ICP-MS）可以高精度地测定盐类中的金属杂质。然而，目前还没有普遍接受的ICP-MS分析盐杂质的方法可以与已建立的金属方法相媲美。在本文中，我们展示了ICP-MS分析（LiF）0.465-(NaF)0.115-(KF)0.42 （FLiNaK）在很大程度上取决于样品制备的每个步骤。探讨了不同的取样方法、盐粒度、溶解酸浓度、消解方法和微波消解条件。讨论了每种方法的最佳实验条件，并对安全性、时间管理和无缝通信的单元选择进行了深入探讨，并通过测量有意添加的金属杂质（MnF2、NiF2、CoF2、FeF2）进行了验证。本研究提出了一种可重复的方法，用于对盐进行精确的ICP-MS测量，以产生更一致和可比较的数据。

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引用次数: 0

ERO2.0 modelling of divertor marker erosion in ASDEX Upgrade L-mode experiments ASDEX Upgrade l型试验中导流器标志侵蚀的ERO2.0模型

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-11-20 DOI: 10.1016/j.nme.2025.102032

S. Saari , A. Hakola , J. Karhunen , M. Balden , C. Baumann , A. Järvinen , K. Krieger , H. Kumpulainen , J. Likonen , J. Romazanov , ASDEX Upgrade Team , EUROfusion Tokamak Exploitation Team

Erosion of small marker surfaces in an experiment conducted at the ASDEX Upgrade tokamak was modelled using the ERO2.0 code. In the experiment 5 × 5 mm² and 1 × 1 mm² Au marker spots were exposed to a series of high-temperature L-mode plasmas in the low-field side strike point region to serve as proxies for measuring net and gross erosion of W, respectively. An ERO2.0 simulation setup was created for the experiment using background plasma produced using OSM and new angle-dependent reflection and sputtering data for Au generated with the SDTrimSP code. The simulated net erosion of the Au markers agreed closely with the measured values. The erosion of the Au markers was induced mainly by the light B, C and N impurities defined as fixed concentrations in the background plasma. The Au markers were found to undergo up to 15–20 times stronger net erosion in comparison to a uniform W surface. This was attributed to 3–4 times stronger gross erosion of Au in comparison to W and deposition of the eroded Au mostly outside of the markers. Consequently, the simulations suggest strongly compromised capability of Au to act as proxy markers for W in erosion studies due to the significantly higher gross erosion yield of Au and insufficient size of the 5 × 5 mm² markers for successful representation of net erosion, as eroded particles migrate along the plasma flow mostly outside the markers.

在ASDEX升级托卡马克上进行的实验中，小标记表面的侵蚀使用ERO2.0代码进行建模。在实验中，将5 × 5 mm2和1 × 1 mm2的Au标记点暴露在低场侧触点区域的一系列高温l模等离子体中，分别作为W的净侵蚀和总侵蚀的测量指标。利用OSM产生的背景等离子体和SDTrimSP代码生成的Au的新角度相关反射和溅射数据，为实验创建了ERO2.0模拟设置。Au标记物的净侵蚀模拟值与实测值吻合较好。Au标记物的侵蚀主要是由背景等离子体中固定浓度的轻B、C和N杂质引起的。与均匀的W表面相比，Au标记物遭受的净侵蚀强度高达15-20倍。这是由于Au的总侵蚀强度是W的3-4倍，并且侵蚀后的Au大多沉积在标记物之外。因此，模拟结果表明，在侵蚀研究中，Au作为W的替代标记物的能力受到严重损害，因为Au的总侵蚀产量明显较高，而且5 × 5 mm2标记物的尺寸不足以成功表示净侵蚀，因为被侵蚀的颗粒主要沿着等离子体流在标记物之外迁移。

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引用次数: 0

Effect of grain refinement on cracks occurring in SUS304L stainless steel under nuclear reactor operating conditions 核反应堆工况下晶粒细化对SUS304L不锈钢裂纹产生的影响

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-10-21 DOI: 10.1016/j.nme.2025.102009

Noriaki Hirota , Ryoma Takeda , Hiroshi ide , Kunihiko Tsuchiya , Yoshinao Kobayashi

Using SUS304L stainless steel, which is employed in reactor structural components, the effects of grain refinement on stress corrosion cracking occurring under nuclear reactor operating conditions were investigated. As a result, after conducting slow strain rate testing (SSRT) in air and nuclear reactor operating environments, a comparison of the tensile properties of SUS304L with the same grain size revealed that elongation significantly decreased with increasing grain size under nuclear reactor operating conditions. In SSRT conducted in air, the

k

-value obtained from the Hall–Petch relationship was lower than the conventional values. Observations showed the absence of cracks on SUS304L with 0.59 and 1.52 µm grains; however, SUS304L with larger grains exhibited rougher fracture surfaces and side cracks. Thin oxide films were formed on SUS304L with 0.59 µm and 1.52 µm grains, while SUS304L with coarse grains of 28.4 µm or larger enabled the formation of oxide films with over 2 µm thickness. Cr₂O₃ films were formed on SUS304L with 0.59 µm, 1.52 µm, and 28.4 µm, while Cr₂O₃ and Fe-based oxides were formed on SUS304L with 39.5 µm and 68.6 µm. Crystal orientation analysis revealed linear surface layers without cracks in the γ-phase for SUS304L with 0.59 µm and 1.52 µm. In materials with larger grain sizes, surface irregularities and cracks were observed in the γ-phase. In fine-grained SUS304L, lattice diffusion caused uniform O diffusion in the γ-phase, resulting in the formation of a thin Cr₂O₃ layer that suppressed cracks. In coarse-grained SUS304L, grain boundary diffusion caused Fe oxide formation at the grain boundaries, weakening them, and supersaturated O led to the formation of thick films comprising Cr₂O₃ and Fe-based oxides, resulting in peeling and cracking.

以SUS304L不锈钢为材料，研究了晶粒细化对核反应堆运行条件下应力腐蚀开裂的影响。因此，在空气和核反应堆运行环境下进行慢应变速率测试（SSRT），对比相同晶粒尺寸SUS304L的拉伸性能发现，在核反应堆运行条件下，伸长率随着晶粒尺寸的增大而显著降低。在空中进行的SSRT中，由Hall-Petch关系得到的k值低于常规值。观察表明，晶粒为0.59和1.52 μ m的SUS304L没有裂纹；而晶粒较大的SUS304L断口表面和侧裂纹较为粗糙。晶粒为0.59µm和1.52µm的SUS304L可以形成较薄的氧化膜，而晶粒为28.4µm及以上的SUS304L可以形成厚度超过2µm的氧化膜。在SUS304L表面形成的Cr2O3薄膜厚度分别为0.59µm、1.52µm和28.4µm，而在SUS304L表面形成的Cr2O3和fe基氧化物厚度分别为39.5µm和68.6µm。晶体取向分析表明，在0.59µm和1.52µm厚度的SUS304L中，γ相为线性面层，无裂纹。在晶粒尺寸较大的材料中，γ相中出现了表面不规则和裂纹。在细晶SUS304L中，晶格扩散导致γ相中O扩散均匀，形成薄的Cr2O3层抑制裂纹。在粗晶SUS304L中，晶界扩散导致晶界处形成Fe氧化物，使晶界变弱，O过饱和导致形成由Cr2O3和Fe基氧化物组成的厚膜，导致剥离和开裂。

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引用次数: 0

Verification, validation, and cross-comparison of tritium transport codes FESTIM, MHIMS, and mHIT 氚输运代码festm、MHIMS和mHIT的验证、确认和交叉比较

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-11-08 DOI: 10.1016/j.nme.2025.102026

Gabriele Ferrero , Raffaella Testoni , Etienne A. Hodille

Tritium transport is a fundamental topic in the development of nuclear fusion reactors for sustainable and competitive energy production. Tritium breeding blankets and extraction systems must be as efficient as possible. Tritium handling systems are crucial to ensure fuel self-sufficiency, safe operations, and cost reduction. Component-level modeling supports design choices to build a more efficient system. In recent years, multiple component-level codes dedicated to simulating hydrogen Isotope transport mechanisms, such as permeation across materials and trapping, have been developed, verified, and validated. This work presents a comparison between three codes, MHIMS, FESTIM, and mHIT, in different verification and validation benchmarks, and their application on the ITER tungsten monoblock. The code comparison includes the V&V study for the mHIT code, and FESTIM results are compared against another code for the ITER monoblock in 2D and during transients. Indeed, to analyze and design tritium components for a fusion power plant, such as a breeder blanket, a plethora of features are necessary, such as trapping, 3 dimensions, multi-material interfaces, time-dependent transients, chemical reactions, and CFD coupling. The benchmarks showcased good agreement between the codes and experimental results. This work demonstrates the coherence and the solid common ground between the codes, verifies some features that are already implemented, and can serve as a starting point for more complex transport features (e.g., chemical reactions, convection, and turbulence coupling).

氚输运是核聚变反应堆可持续和有竞争力的能源生产发展的一个基本问题。氚繁殖毯和提取系统必须尽可能高效。氚处理系统对于确保燃料自给自足、安全运行和降低成本至关重要。组件级建模支持设计选择，以构建更高效的系统。近年来，多个组件级代码致力于模拟氢同位素传输机制，如跨材料渗透和捕获，已经开发，验证和验证。本文介绍了MHIMS、festm和mHIT三种代码在不同验证基准中的比较，以及它们在ITER钨块上的应用。代码比较包括对mHIT代码的V&；V研究，并将festm结果与ITER单块的另一个代码在二维和瞬态期间进行比较。事实上，为了分析和设计核聚变发电厂的氚组件，如增殖毯，大量的特征是必要的，如捕获、三维、多材料界面、随时间变化的瞬态、化学反应和CFD耦合。基准测试表明，代码与实验结果吻合良好。这项工作展示了代码之间的一致性和坚实的共同点，验证了一些已经实现的特征，并且可以作为更复杂的传输特征（例如，化学反应，对流和湍流耦合）的起点。

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引用次数: 0

Thermophysical properties and expansion anisotropy of sintered hafnium hydride compacts 烧结氢化铪致密体的热物理性质和膨胀各向异性

IF 2.7 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2025-12-01 Epub Date: 2025-10-24 DOI: 10.1016/j.nme.2025.102014

J.P. Pollard , T. Zagyva , C.S.J. Pickles , J.O. Astbury , C.G. Windsor , A. Shivprasad , C.A. Kohnert , F. Giuliani , S. Humphry-Baker

Thermophysical properties are reported on ε-HfH₂ samples fabricated by powder metallurgy. Samples were heat treated in the range 300–550 °C to transform them from ε-HfH₂ to δ-HfH_1.6-x, allowing comparison of the properties of both phases. Higher molar heat capacity was found in stoichiometric ε-HfH₂ compared to literature data on sub-stoichiometric ε-HfH_1.83. The δ-phase undergoes a vacancy order–disorder transformation at ∼130 °C with a transformation enthalpy of ∼1.4 kJ mol⁻¹. The room-temperature thermal diffusivity of the ε and δ phases were 0.11 and 0.09 cm² s⁻¹ respectively. These values are lower than those for literature bulk hydride materials, which is accounted for by pore-phonon scattering. Thermal expansion of ε and δ phases was measured by high-temperature X-ray diffraction to be 9.2 and 11 x10^-6 K⁻¹, respectively. The data on the ε phase is the first known in the literature. The thermal expansion was highly anisotropic, with a negative thermal expansion parallel to the a-axis (R_a = −8.7). Such extreme anisotropy has implications in controlling the microstructure for thermal damage tolerance.

报道了粉末冶金法制备的ε-HfH2样品的热物理性质。在300-550℃范围内对样品进行热处理，使其从ε-HfH2转变为δ-HfH1.6-x，从而比较两相的性能。与文献中亚化学计量量ε-HfH2的数据相比，发现其摩尔热容更高。δ相在~ 130℃发生空位有序-无序转变，转变焓为~ 1.4 kJ mol−1。ε相和δ相的室温热扩散系数分别为0.11和0.09 cm2 s−1。这些数值低于文献中块状氢化物材料的数值，这是由孔声子散射引起的。高温x射线衍射测得ε相和δ相的热膨胀分别为9.2和11 x10-6 K−1。ε相的数据是文献中已知的第一个。热膨胀具有高度的各向异性，负热膨胀平行于a轴（Ra = - 8.7）。这种极端的各向异性对控制热损伤容限的微观结构具有重要意义。

{"title":"Thermophysical properties and expansion anisotropy of sintered hafnium hydride compacts","authors":"J.P. Pollard , T. Zagyva , C.S.J. Pickles , J.O. Astbury , C.G. Windsor , A. Shivprasad , C.A. Kohnert , F. Giuliani , S. Humphry-Baker","doi":"10.1016/j.nme.2025.102014","DOIUrl":"10.1016/j.nme.2025.102014","url":null,"abstract":"<div><div>Thermophysical properties are reported on ε-HfH<sub>2</sub> samples fabricated by powder metallurgy. Samples were heat treated in the range 300–550 °C to transform them from ε-HfH<sub>2</sub> to δ-HfH<sub>1.6-x</sub>, allowing comparison of the properties of both phases. Higher molar heat capacity was found in stoichiometric ε-HfH<sub>2</sub> compared to literature data on sub-stoichiometric ε-HfH<sub>1.83</sub>. The δ-phase undergoes a vacancy order–disorder transformation at ∼130 °C with a transformation enthalpy of ∼1.4 kJ mol<sup>−1</sup>. The room-temperature thermal diffusivity of the ε and δ phases were 0.11 and 0.09 cm<sup>2</sup> s<sup>−1</sup> respectively. These values are lower than those for literature bulk hydride materials, which is accounted for by pore-phonon scattering. Thermal expansion of ε and δ phases was measured by high-temperature X-ray diffraction to be 9.2 and 11 x10<sup>-6</sup> K<sup>−1</sup>, respectively. The data on the ε phase is the first known in the literature. The thermal expansion was highly anisotropic, with a negative thermal expansion parallel to the a-axis (R<sub>a</sub> = −8.7). Such extreme anisotropy has implications in controlling the microstructure for thermal damage tolerance.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102014"},"PeriodicalIF":2.7,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145417636","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

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