Nuclear Materials and Energy最新文献_第4页

Promising cooling concepts for enhanced JT-60SA tungsten actively cooled divertor 有前途的冷却概念，增强型JT-60SA钨主动冷却分流器

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

Nuclear Materials and Energy

Pub Date : 2025-11-19 DOI: 10.1016/j.nme.2025.102030

D. Dias Aleixo , M. Firdaouss , T. Baffie , P.-E. Frayssines , H. Gleyzes , P. Lechevalier , H. Roche , E. Tejado , A. Thomas , V. Tomarchio , M. Richou

To propose enhanced concepts for the JT-60SA tungsten target divertor able to withstand heat loads higher than 20 MW/m², this paper investigates the flat tile design (flat junction between tungsten armour material and heat sink) using additively manufactured CuCrZr heat sinks. Two enhanced hypervapotron cooling channel designs, called in this paper HV Diagonal and HV Chevron, to efficiently cool the heat sink, are investigated. The components are produced via Laser Powder Bed Fusion (LPBF) and post-processed by Hot Isostatic Pressing (HIP) to close the residual pores coming from the additive manufacturing technique and to simulate the diffusion bonding between the heat sink and tungsten. Computational Fluid Dynamics (CFD) analysis shows that HV Chevron and HV Diagonal designs are promising, as they result in lower inner wall temperatures, up to 40 °C lower at an incident heat flux of 7 MW/m², compared to the conventional hypervapotron design, with an increase of pressure drop about 30 %. These findings are supported by High Heat Flux (HHF) tests, where both mock-ups withstood a heat flux of up to 20 MW/m² for HV Diagonal and 25 MW/m² for HV Chevron, both in steady-state regime. Preliminary results reveal that LPBF CuCrZr requires a water quench thermal treatment to meet the CuCrZr mechanical properties of the ITER specifications. After thermal treatments, the LPBF CuCrZr material reached a relative density of 99.6 %, with all initial pores effectively closed. This paper presents the potential of the combination of CFD simulation and additive manufacturing for plasma-facing components and demonstrates, as a first step, the feasibility of using LPBF combined with the HIP process for the fabrication of plasma-facing components using CuCrZr as heat sink.

为了提出能够承受高于20 MW/m2热负荷的JT-60SA钨靶导向器的增强概念，本文研究了使用增材制造的CuCrZr散热器的平瓦设计（钨装甲材料和散热器之间的平结）。本文研究了两种增强型超蒸汽冷却通道设计，即HV对角线和HV雪佛龙，以有效地冷却散热器。通过激光粉末床熔合（LPBF）和热等静压（HIP）后处理来关闭增材制造技术产生的残余孔隙，并模拟散热器与钨之间的扩散结合。计算流体动力学（CFD）分析表明，与传统的超蒸汽设计相比，HV Chevron和HV Diagonal设计具有较低的内壁温度，在入射热流密度为7 MW/m2时，内壁温度可降低40°C，压降提高约30%。这些发现得到了高热流密度（HHF）测试的支持，两种模型在稳态状态下都能承受高达20 MW/m2的HV对角和25 MW/m2的HV雪佛龙的热流密度。初步结果表明，LPBF CuCrZr需要水淬热处理才能满足ITER规范的CuCrZr力学性能。热处理后，LPBF CuCrZr材料的相对密度达到99.6%，初始孔隙全部有效闭合。本文介绍了将CFD模拟与增材制造相结合用于等离子体表面部件的潜力，并作为第一步，论证了将LPBF与HIP工艺相结合用于以CuCrZr为散热器制造等离子体表面部件的可行性。

{"title":"Promising cooling concepts for enhanced JT-60SA tungsten actively cooled divertor","authors":"D. Dias Aleixo , M. Firdaouss , T. Baffie , P.-E. Frayssines , H. Gleyzes , P. Lechevalier , H. Roche , E. Tejado , A. Thomas , V. Tomarchio , M. Richou","doi":"10.1016/j.nme.2025.102030","DOIUrl":"10.1016/j.nme.2025.102030","url":null,"abstract":"<div><div>To propose enhanced concepts for the JT-60SA tungsten target divertor able to withstand heat loads higher than 20 MW/m<sup>2</sup>, this paper investigates the flat tile design (flat junction between tungsten armour material and heat sink) using additively manufactured CuCrZr heat sinks. Two enhanced hypervapotron cooling channel designs, called in this paper HV Diagonal and HV Chevron, to efficiently cool the heat sink, are investigated. The components are produced via Laser Powder Bed Fusion (LPBF) and post-processed by Hot Isostatic Pressing (HIP) to close the residual pores coming from the additive manufacturing technique and to simulate the diffusion bonding between the heat sink and tungsten. Computational Fluid Dynamics (CFD) analysis shows that HV Chevron and HV Diagonal designs are promising, as they result in lower inner wall temperatures, up to 40 °C lower at an incident heat flux of 7 MW/m<sup>2</sup>, compared to the conventional hypervapotron design, with an increase of pressure drop about 30 %. These findings are supported by High Heat Flux (HHF) tests, where both mock-ups withstood a heat flux of up to 20 MW/m<sup>2</sup> for HV Diagonal and 25 MW/m<sup>2</sup> for HV Chevron, both in steady-state regime. Preliminary results reveal that LPBF CuCrZr requires a water quench thermal treatment to meet the CuCrZr mechanical properties of the ITER specifications. After thermal treatments, the LPBF CuCrZr material reached a relative density of 99.6 %, with all initial pores effectively closed. This paper presents the potential of the combination of CFD simulation and additive manufacturing for plasma-facing components and demonstrates, as a first step, the feasibility of using LPBF combined with the HIP process for the fabrication of plasma-facing components using CuCrZr as heat sink.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102030"},"PeriodicalIF":2.7,"publicationDate":"2025-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145579287","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}

引用次数: 0

Boron layer preparation, characterization and hydrogen isotope permeability for fusion application 硼层制备、表征及聚变应用中氢同位素渗透率

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

Nuclear Materials and Energy

Pub Date : 2025-11-15 DOI: 10.1016/j.nme.2025.102028

A. Houben , E. Warkentin , M. Rasiński , T. Dittmar , H.R. Koslowski , S. Möller , B. Unterberg , Ch. Linsmeier

Due to the re-baseline of the fusion device ITER and the strategical decision to change from Be to W as first wall material, a boronization procedure has to be implemented into the wall conditioning phase. Since the functionality of boron layers in carbon free fusion devices is not understand in detail so far, this study aims to be a starting point of the investigation of boron layers for fusion applications.

In the first step, pure boron coatings are prepared in a magnetron sputter deposition device on W and steel substrates. The homogeneity, crystal phase and composition is studied and it is proved that an amorphous, stable boron layer is obtained with this deposition procedure. No impurities, e.g. O, N, C, are detected and a deposition rate of 20 nm/h is reached. The coatings are temperature stable up to 1000

^{\circ} C

. No oxidation of the boron layer is detected when exposed to air, but a uptake of humidity is possible. Therefore, the samples should be stored in vacuum after deposition.

The hydrogen isotope permeability is studied and a low layer permeability, which is four orders of magnitude lower as steel is found.

In the future, the investigation will be broadened to mixed boron layers, e.g. B:D and B:W, which are more alike as boron layers in fusion devices, and these mixed layers will be compared to the pure boron layers as a next step.

由于核聚变装置ITER的重新基线以及从Be改为W作为第一壁材的战略决策，必须在壁材调节阶段实施硼化程序。由于目前对无碳聚变装置中硼层的功能尚不了解，本研究旨在成为硼层聚变应用研究的一个起点。在第一步中，在磁控溅射沉积装置中在W和钢基体上制备纯硼涂层。研究了硼的均匀性、晶相和组成，证明了该沉积工艺可获得稳定的非晶硼层。未检测到O、N、C等杂质，沉积速率可达20 nm/h。这种涂层在1000°C下温度稳定。当暴露在空气中时，硼层不会被氧化，但可能会吸收湿度。因此，沉积后的样品应真空保存。对氢同位素渗透率进行了研究，发现低层渗透率比钢低4个数量级。未来，研究将扩展到混合硼层，如B:D和B:W，它们更像聚变装置中的硼层，并将这些混合层与纯硼层进行比较。

{"title":"Boron layer preparation, characterization and hydrogen isotope permeability for fusion application","authors":"A. Houben , E. Warkentin , M. Rasiński , T. Dittmar , H.R. Koslowski , S. Möller , B. Unterberg , Ch. Linsmeier","doi":"10.1016/j.nme.2025.102028","DOIUrl":"10.1016/j.nme.2025.102028","url":null,"abstract":"<div><div>Due to the re-baseline of the fusion device ITER and the strategical decision to change from Be to W as first wall material, a boronization procedure has to be implemented into the wall conditioning phase. Since the functionality of boron layers in carbon free fusion devices is not understand in detail so far, this study aims to be a starting point of the investigation of boron layers for fusion applications.</div><div>In the first step, pure boron coatings are prepared in a magnetron sputter deposition device on W and steel substrates. The homogeneity, crystal phase and composition is studied and it is proved that an amorphous, stable boron layer is obtained with this deposition procedure. No impurities, e.g. O, N, C, are detected and a deposition rate of 20 nm/h is reached. The coatings are temperature stable up to 1000 <span><math><mrow><msup><mrow></mrow><mrow><mo>∘</mo></mrow></msup><mtext>C</mtext></mrow></math></span>. No oxidation of the boron layer is detected when exposed to air, but a uptake of humidity is possible. Therefore, the samples should be stored in vacuum after deposition.</div><div>The hydrogen isotope permeability is studied and a low layer permeability, which is four orders of magnitude lower as steel is found.</div><div>In the future, the investigation will be broadened to mixed boron layers, e.g. B:D and B:W, which are more alike as boron layers in fusion devices, and these mixed layers will be compared to the pure boron layers as a next step.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102028"},"PeriodicalIF":2.7,"publicationDate":"2025-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578616","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}

引用次数: 0

Dispersoid coarsening induced softening in Ti-doped ODS-Cu alloys under Fe ion irradiation at 350°C 350℃Fe辐照下ti掺杂ODS-Cu合金分散体粗化引起的软化

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

Nuclear Materials and Energy

Pub Date : 2025-11-15 DOI: 10.1016/j.nme.2025.102027

Sixiang Zhao , Guowei Song , Yusheng Zhang , Yuheng Zhang , Chonghong Zhang , Guangnan Luo

Oxide dispersion strengthened copper (ODS-Cu) alloy is considered as a candidate of heat sink materials for divertors, and Ti doping is regarded as an effective strategy to improve the performance of Cu-Al₂O₃ (a common ODS-Cu). In order to evaluate the stability of the oxide nano-particles dispersed in Ti-doped ODS-Cu, irradiation was conducted using multiple-energy Fe ions at 350°C on two types of Ti-doped ODS-Cu with identical composition and oxide volumetric fraction while different size of oxide nano-particles. Ultimately, a 24 μm-thick quasi-homogeneous damaged layer of ∼ 1.35 dpa was induced in the specimens. Nano-hardness measurements were used to assess variation in the mechanical properties of the irradiated ODS-Cu. The results showed that both specimens experienced irradiation-induced softening, and softening in the specimen containing smaller sized particles is more pronounced. Microscopic observations reveal that the size of irradiated oxide nano-particles somewhat increased, which can be mainly explained by Ostwald ripening under irradiation, and this ripening effect is more pronounced in the smaller oxide nano-particles.

氧化物弥散强化铜（ODS-Cu）合金是一种新型的热沉材料，Ti掺杂是提高Cu-Al2O3（一种常见的ODS-Cu）性能的有效策略。为了评价分散在ti掺杂ODS-Cu中的氧化物纳米颗粒的稳定性，采用多能Fe离子在350°C下对两种成分和氧化物体积分数相同、氧化物纳米颗粒尺寸不同的ti掺杂ODS-Cu进行辐照。最终，在样品中诱导出24 μm厚的准均匀损伤层，厚度约为1.35 dpa。采用纳米硬度测量来评估辐照后ODS-Cu的力学性能变化。结果表明：两种试样均经历辐照诱导软化，且颗粒较小的试样软化更为明显；微观观察表明，辐照后的氧化纳米颗粒的尺寸有所增大，这主要是由于辐照下的奥斯特瓦尔德成熟，并且这种成熟效应在较小的氧化纳米颗粒中更为明显。

{"title":"Dispersoid coarsening induced softening in Ti-doped ODS-Cu alloys under Fe ion irradiation at 350°C","authors":"Sixiang Zhao , Guowei Song , Yusheng Zhang , Yuheng Zhang , Chonghong Zhang , Guangnan Luo","doi":"10.1016/j.nme.2025.102027","DOIUrl":"10.1016/j.nme.2025.102027","url":null,"abstract":"<div><div>Oxide dispersion strengthened copper (ODS-Cu) alloy is considered as a candidate of heat sink materials for divertors, and Ti doping is regarded as an effective strategy to improve the performance of Cu-Al<sub>2</sub>O<sub>3</sub> (a common ODS-Cu). In order to evaluate the stability of the oxide nano-particles dispersed in Ti-doped ODS-Cu, irradiation was conducted using multiple-energy Fe ions at 350°C on two types of Ti-doped ODS-Cu with identical composition and oxide volumetric fraction while different size of oxide nano-particles. Ultimately, a 24 μm-thick quasi-homogeneous damaged layer of ∼ 1.35 dpa was induced in the specimens. Nano-hardness measurements were used to assess variation in the mechanical properties of the irradiated ODS-Cu. The results showed that both specimens experienced irradiation-induced softening, and softening in the specimen containing smaller sized particles is more pronounced. Microscopic observations reveal that the size of irradiated oxide nano-particles somewhat increased, which can be mainly explained by Ostwald ripening under irradiation, and this ripening effect is more pronounced in the smaller oxide nano-particles.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102027"},"PeriodicalIF":2.7,"publicationDate":"2025-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578617","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}

引用次数: 0

Hydrogen ion-induced surface damage of copper grids in RF ion sources for fusion NBI 聚变NBI射频离子源中氢离子诱导铜网表面损伤

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

Nuclear Materials and Energy

Pub Date : 2025-11-12 DOI: 10.1016/j.nme.2025.102013

Xiaona Li , Jianhua Lv , Weiyuan Ni , Chao Chen , Miao Zhao , Xingquan Wang , Mengchao Li , Guangjiu Lei

During the operation of extracting hydrogen ions in NBI RF ion systems, the plasma grid is exposed to prolonged irradiation with low-energy hydrogen ions, leading to surface damage. In this study, a hydrogen plasma environment is constructed to investigate hydrogen ion-induced surface damage of copper grid. During the irradiation, the ion extraction process is considered. Additionally, a numerical fluid model was developed to analyze ion implantation parameters on the flat and conical surfaces of the extraction aperture under different gas pressures and extraction voltages. The results reveal that different extraction voltages influence the electric field in the extraction region, thereby affecting the energy of hydrogen ion implantation and resulting in surface damage. Surprisingly, the copper grid exhibits more severe surface swelling on the conical surface compared to the flat surface after prolonged irradiation, even though the irradiation flux on the conical surface is approximately half that on the flat surface. The behavior of severe swelling on the conical surface can be attributed to the synergistic effect of suppression and shadowing. The surface swelling induced by hydrogen ions can alter surface roughness, subsequently impacting work function and the efficiency and stability of ion beam extraction.

在NBI射频离子系统中提取氢离子的过程中，等离子体栅格暴露在低能氢离子的长时间照射下，导致表面损伤。在本研究中，我们构建了一个氢等离子体环境来研究氢离子对铜栅极表面的损伤。在辐照过程中，考虑了离子萃取过程。此外，建立了数值流体模型，分析了不同气体压力和提取电压下离子在提取孔平面和锥形表面的注入参数。结果表明，不同的萃取电压会影响萃取区电场，从而影响氢离子注入能量，导致表面损伤。令人惊讶的是，尽管锥形表面的辐照通量约为平面表面的一半，但与平面相比，锥形表面的铜网格在长时间照射后表现出更严重的表面膨胀。锥形表面的剧烈膨胀行为可归因于抑制和遮蔽的协同作用。氢离子引起的表面膨胀会改变表面粗糙度，进而影响功函数和离子束萃取的效率和稳定性。

{"title":"Hydrogen ion-induced surface damage of copper grids in RF ion sources for fusion NBI","authors":"Xiaona Li , Jianhua Lv , Weiyuan Ni , Chao Chen , Miao Zhao , Xingquan Wang , Mengchao Li , Guangjiu Lei","doi":"10.1016/j.nme.2025.102013","DOIUrl":"10.1016/j.nme.2025.102013","url":null,"abstract":"<div><div>During the operation of extracting hydrogen ions in NBI RF ion systems, the plasma grid is exposed to prolonged irradiation with low-energy hydrogen ions, leading to surface damage. In this study, a hydrogen plasma environment is constructed to investigate hydrogen ion-induced surface damage of copper grid. During the irradiation, the ion extraction process is considered. Additionally, a numerical fluid model was developed to analyze ion implantation parameters on the flat and conical surfaces of the extraction aperture under different gas pressures and extraction voltages. The results reveal that different extraction voltages influence the electric field in the extraction region, thereby affecting the energy of hydrogen ion implantation and resulting in surface damage. Surprisingly, the copper grid exhibits more severe surface swelling on the conical surface compared to the flat surface after prolonged irradiation, even though the irradiation flux on the conical surface is approximately half that on the flat surface. The behavior of severe swelling on the conical surface can be attributed to the synergistic effect of suppression and shadowing. The surface swelling induced by hydrogen ions can alter surface roughness, subsequently impacting work function and the efficiency and stability of ion beam extraction.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102013"},"PeriodicalIF":2.7,"publicationDate":"2025-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578613","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}

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 : 2025-11-10 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。

{"title":"Role of doped ZrC on deuterium trapping in W-ZrC alloy","authors":"Xuexi Zhang , Li Qiao , Hong Zhang , Xuefeng Xie , Yange Zhang , Peng Wang , Changsong Liu","doi":"10.1016/j.nme.2025.102023","DOIUrl":"10.1016/j.nme.2025.102023","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102023"},"PeriodicalIF":2.7,"publicationDate":"2025-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578615","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}

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 : 2025-11-08 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。

{"title":"Verification, validation, and cross-comparison of tritium transport codes FESTIM, MHIMS, and mHIT","authors":"Gabriele Ferrero , Raffaella Testoni , Etienne A. Hodille","doi":"10.1016/j.nme.2025.102026","DOIUrl":"10.1016/j.nme.2025.102026","url":null,"abstract":"<div><div>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).</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102026"},"PeriodicalIF":2.7,"publicationDate":"2025-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145528698","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}

引用次数: 0

Helium bubbles retard recrystallization in tungsten by limiting subgrain growth 氦气泡通过限制亚晶生长来延缓钨的再结晶

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

Nuclear Materials and Energy

Pub Date : 2025-11-07 DOI: 10.1016/j.nme.2025.102024

Chang Xu , Yu Li , Yiwen Zhu , Yu Tian , Jing Liang , Longqiang Han , Zutao Guo , Qiang Li , Guang-Nan Luo , Hai-Shan Zhou

The tungsten plasma facing components (PFCs) in fusion devices must withstand high heat loads. Recrystallization sets in under such conditions and embrittles tungsten, hence motivating its mitigation. He plasma exposure, a natural consequence of the deuterium–tritium reaction, has been shown to retard recrystallization, the mechanism of which was generally attributed to the Zener pinning of grain boundaries during the grain growth stage of recrystallization. However, neither the Zener pinning effect was quantified nor its role in the nucleation stage of recrystallization was considered, limiting our understanding and exploitation of this beneficial phenomenon. Here, we approach this problem by splitting the conventional annealing experiment into finer steps using a powerful electron beam at 1573 K. After each 15 min annealing of the He plasma exposed tungsten specimen, the same area was examined by electron backscatter diffraction (EBSD) to detect recrystallized grains and transmission electron microscopy (TEM) to observe subsurface He bubbles. We found that near-surface recrystallization was retarded at 1573 K. Surprisingly, site-specific TEM imaging revealed that recrystallization was retarded even when the grain boundaries of the recrystallized grains were not pinned by He bubbles, which directly contradicts previous assumptions. Additionally, the pressure balance on the grain boundaries suggested that grain growth was unfavorable, even in the absence of the Zener pinning pressure. Instead, recrystallization and its suppression were dominated by nucleation, as supported by EBSD analysis of the nucleation rate. A model based on He bubbles limited subgrain growth was developed and quantitatively agreed with the recrystallization fraction trend at 1573 K. Overall, we demonstrate He bubbles limited subgrain growth as a viable mechanism of retarding recrystallization. Moreover, it is a promising model for identifying a regime of retarded recrystallization for tungsten PFCs in fusion devices.

聚变装置中的钨等离子体表面元件（pfc）必须承受高热负荷。在这种条件下，再结晶形成并使钨变脆，从而促使钨的软化。氦等离子体暴露是氘-氚反应的自然结果，已被证明可以延缓再结晶，其机制通常归因于再结晶晶粒生长阶段晶界的齐纳钉扎。然而，既没有量化齐纳钉钉效应，也没有考虑其在再结晶成核阶段的作用，限制了我们对这一有益现象的理解和利用。在这里，我们通过使用1573 K的强大电子束将传统的退火实验分成更精细的步骤来解决这个问题。在He等离子体暴露的钨样品每15 min退火后，用电子背散射衍射（EBSD）检测同一区域的再结晶颗粒，用透射电子显微镜（TEM）观察表面下He气泡。在1573 K时，近表面再结晶被延缓。令人惊讶的是，特定位置的TEM成像显示，即使在再结晶晶粒的晶界没有被He气泡固定的情况下，再结晶也会被延缓，这直接与之前的假设相矛盾。此外，晶界上的压力平衡表明，即使在没有齐纳钉压的情况下，晶粒的生长也是不利的。相反，再结晶和再结晶的抑制主要是成核，这一点得到了成核速率的EBSD分析的支持。建立了基于He气泡限制亚晶生长的模型，该模型与1573 K时的再结晶分数趋势相吻合。总之，我们证明了He气泡限制亚晶生长是一种可行的延缓再结晶的机制。此外，它是一个很有前途的模型，用于确定熔合装置中钨PFCs的延迟再结晶状态。

{"title":"Helium bubbles retard recrystallization in tungsten by limiting subgrain growth","authors":"Chang Xu , Yu Li , Yiwen Zhu , Yu Tian , Jing Liang , Longqiang Han , Zutao Guo , Qiang Li , Guang-Nan Luo , Hai-Shan Zhou","doi":"10.1016/j.nme.2025.102024","DOIUrl":"10.1016/j.nme.2025.102024","url":null,"abstract":"<div><div>The tungsten plasma facing components (PFCs) in fusion devices must withstand high heat loads. Recrystallization sets in under such conditions and embrittles tungsten, hence motivating its mitigation. He plasma exposure, a natural consequence of the deuterium–tritium reaction, has been shown to retard recrystallization, the mechanism of which was generally attributed to the Zener pinning of grain boundaries during the grain growth stage of recrystallization. However, neither the Zener pinning effect was quantified nor its role in the nucleation stage of recrystallization was considered, limiting our understanding and exploitation of this beneficial phenomenon. Here, we approach this problem by splitting the conventional annealing experiment into finer steps using a powerful electron beam at 1573 K. After each 15 min annealing of the He plasma exposed tungsten specimen, the same area was examined by electron backscatter diffraction (EBSD) to detect recrystallized grains and transmission electron microscopy (TEM) to observe subsurface He bubbles. We found that near-surface recrystallization was retarded at 1573 K. Surprisingly, site-specific TEM imaging revealed that recrystallization was retarded even when the grain boundaries of the recrystallized grains were not pinned by He bubbles, which directly contradicts previous assumptions. Additionally, the pressure balance on the grain boundaries suggested that grain growth was unfavorable, even in the absence of the Zener pinning pressure. Instead, recrystallization and its suppression were dominated by nucleation, as supported by EBSD analysis of the nucleation rate. A model based on He bubbles limited subgrain growth was developed and quantitatively agreed with the recrystallization fraction trend at 1573 K. Overall, we demonstrate He bubbles limited subgrain growth as a viable mechanism of retarding recrystallization. Moreover, it is a promising model for identifying a regime of retarded recrystallization for tungsten PFCs in fusion devices.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102024"},"PeriodicalIF":2.7,"publicationDate":"2025-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145528700","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}

引用次数: 0

Molecular dynamics simulation of tungsten surface sputtering under low–energy neon irradiation for radiative divertor scenarios 低能氖辐射散射下钨表面溅射的分子动力学模拟

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

Nuclear Materials and Energy

Pub Date : 2025-11-07 DOI: 10.1016/j.nme.2025.102025

Xianglong Liu , Zhongshi Yang , Guojian Niu , Xiaochun Li , Yilang Mai , Jing Liang , Yilin Zhou , Kedong Li , Haishan Zhou , Guang-Nan Luo

Neon (Ne) is a key candidate impurity for radiative divertor operation in tokamaks, aiming to reduce the thermal load on the divertor targets. However, its injection can simultaneously enhance tungsten surface erosion. In this work, molecular dynamics simulations are employed to investigate the sputtering behavior of tungsten under Ne irradiation in the energy range of 30–200 eV. The effects of incident energy, crystallographic orientation, and incidence angle are systematically analyzed, and the evolution of surface morphology under cumulative bombardment is examined. The results indicate that the sputtering yield increases significantly with incident energy, with the (110) surface exhibiting the highest yield due to its high atomic density and shallow Ne implantation depth. Sputtering yields are further enhanced at specific oblique incidence angles due to more efficient momentum transfer to surface atoms. The angular and energy distributions of sputtered atoms show that low–energy sputtering is more sensitive to surface orientation. Under cumulative injection, Ne retention increases and saturates with fluence, while surface damage strongly depends on the irradiation energy and fluence. Furthermore, the growth of surface roughness leads to a reduction in sputtering yield. These findings provide microscopic insight into impurity–tungsten interactions in fusion devices and offer quantitative input for boundary plasma modeling and divertor lifetime assessment.

氖（Ne）是托卡马克辐射导流器运行的关键候选杂质，旨在降低导流器靶材的热负荷。但它的注入会同时加剧钨的表面侵蚀。本文采用分子动力学模拟方法研究了钨在30 ~ 200 eV能量范围内的Ne辐照溅射行为。系统地分析了入射能量、晶体取向和入射角的影响，并研究了累积轰击下表面形貌的演变。结果表明，随着入射能量的增加，溅射产率显著增加，其中（110）表面由于其高原子密度和较浅的Ne注入深度而表现出最高的产率。在特定的斜入射角下，由于更有效的动量传递到表面原子，溅射收率进一步提高。溅射原子的角分布和能量分布表明，低能溅射对表面取向更为敏感。累积注入时，Ne的滞留量随辐照能量的增加而增加并趋于饱和，而表面损伤则强烈依赖于辐照能量和辐照通量。此外，表面粗糙度的增加导致溅射成品率的降低。这些发现提供了对聚变装置中杂质-钨相互作用的微观洞察，并为边界等离子体建模和分流器寿命评估提供了定量输入。

{"title":"Molecular dynamics simulation of tungsten surface sputtering under low–energy neon irradiation for radiative divertor scenarios","authors":"Xianglong Liu , Zhongshi Yang , Guojian Niu , Xiaochun Li , Yilang Mai , Jing Liang , Yilin Zhou , Kedong Li , Haishan Zhou , Guang-Nan Luo","doi":"10.1016/j.nme.2025.102025","DOIUrl":"10.1016/j.nme.2025.102025","url":null,"abstract":"<div><div>Neon (Ne) is a key candidate impurity for radiative divertor operation in tokamaks, aiming to reduce the thermal load on the divertor targets. However, its injection can simultaneously enhance tungsten surface erosion. In this work, molecular dynamics simulations are employed to investigate the sputtering behavior of tungsten under Ne irradiation in the energy range of 30–200 eV. The effects of incident energy, crystallographic orientation, and incidence angle are systematically analyzed, and the evolution of surface morphology under cumulative bombardment is examined. The results indicate that the sputtering yield increases significantly with incident energy, with the (110) surface exhibiting the highest yield due to its high atomic density and shallow Ne implantation depth. Sputtering yields are further enhanced at specific oblique incidence angles due to more efficient momentum transfer to surface atoms. The angular and energy distributions of sputtered atoms show that low–energy sputtering is more sensitive to surface orientation. Under cumulative injection, Ne retention increases and saturates with fluence, while surface damage strongly depends on the irradiation energy and fluence. Furthermore, the growth of surface roughness leads to a reduction in sputtering yield. These findings provide microscopic insight into impurity–tungsten interactions in fusion devices and offer quantitative input for boundary plasma modeling and divertor lifetime assessment.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102025"},"PeriodicalIF":2.7,"publicationDate":"2025-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145528701","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}

引用次数: 0

First Demonstration of laser induced breakdown spectroscopy using remote handling for in-vessel analysis of JET components 首次演示使用远程处理的激光诱导击穿光谱用于JET组件的容器内分析

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

Nuclear Materials and Energy

Pub Date : 2025-11-05 DOI: 10.1016/j.nme.2025.102021

J. Likonen , S. Almaviva , R. Rayaprolu , R. Yi , I. Jepu , G. Sergienko , A. Widdowson , N. Jones , S. Atikukke , T. Dittmar , J. Karhunen , P. Gasior , Ch. Kawan , M. Sackers , S. Soni , E. Wüst , S. Brezinsek , J. Butikova , W. Gromelski , A. Hakola , P. Veis

The feasibility of laser-induced breakdown spectroscopy (LIBS) for measuring fuel retention was demonstrated for the first time in a tokamak operating with tritium using a remotely controlled in-situ application in JET. In JET as well as in future fusion reactors such as ITER and DEMO, thick co-deposited layers will be formed on the inner wall during extended plasma operations. Experiments in present-day fusion devices indicate that these layers consist of eroded plasma-facing materials, various impurities and plasma fuel species such as deuterium and tritium. Accumulation of radioactive tritium in the reactor vacuum vessel is a particularly critical safety issue requiring active monitoring. LIBS is one of the few techniques available for monitoring the tritium content and the composition of co-deposited layers during maintenance breaks [1]. This paper will provide an overview of the LIBS experiment that was performed post DTE3 campaign, and D and H cleanup at JET in October 2024. 840 different spatial locations on the main wall and divertor area were investigated, with measurements of more than 100 laser pulses per location executed to ensure good depth resolution.

激光诱导击穿光谱（LIBS）用于测量燃料保留的可行性首次在带有氚的托卡马克上进行了验证。在JET以及未来的核聚变反应堆（如ITER和DEMO）中，在延长等离子体操作期间，将在内壁形成厚的共沉积层。目前核聚变装置的实验表明，这些层由侵蚀的等离子体表面材料、各种杂质和等离子体燃料物质（如氘和氚）组成。放射性氚在反应堆真空容器中的积累是一个特别关键的安全问题，需要积极监测。LIBS是为数不多的可用于监测氚含量和共沉积层组成的技术之一。本文将概述2024年10月在JET进行的DTE3战役和D和H清理后进行的LIBS实验。研究人员调查了主壁和导流器区域840个不同的空间位置，每个位置测量了100多个激光脉冲，以确保良好的深度分辨率。

{"title":"First Demonstration of laser induced breakdown spectroscopy using remote handling for in-vessel analysis of JET components","authors":"J. Likonen , S. Almaviva , R. Rayaprolu , R. Yi , I. Jepu , G. Sergienko , A. Widdowson , N. Jones , S. Atikukke , T. Dittmar , J. Karhunen , P. Gasior , Ch. Kawan , M. Sackers , S. Soni , E. Wüst , S. Brezinsek , J. Butikova , W. Gromelski , A. Hakola , P. Veis","doi":"10.1016/j.nme.2025.102021","DOIUrl":"10.1016/j.nme.2025.102021","url":null,"abstract":"<div><div>The feasibility of laser-induced breakdown spectroscopy (LIBS) for measuring fuel retention was demonstrated for the first time in a tokamak operating with tritium using a remotely controlled in-situ application in JET. In JET as well as in future fusion reactors such as ITER and DEMO, thick co-deposited layers will be formed on the inner wall during extended plasma operations. Experiments in present-day fusion devices indicate that these layers consist of eroded plasma-facing materials, various impurities and plasma fuel species such as deuterium and tritium. Accumulation of radioactive tritium in the reactor vacuum vessel is a particularly critical safety issue requiring active monitoring. LIBS is one of the few techniques available for monitoring the tritium content and the composition of co-deposited layers during maintenance breaks [<span><span>1</span></span>]. This paper will provide an overview of the LIBS experiment that was performed post DTE3 campaign, and D and H cleanup at JET in October 2024. 840 different spatial locations on the main wall and divertor area were investigated, with measurements of more than 100 laser pulses per location executed to ensure good depth resolution.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102021"},"PeriodicalIF":2.7,"publicationDate":"2025-11-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145474207","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}

引用次数: 0

Water-cooled lead and ceramic breeder (WLCB) breeding blanket (BB) for the EU DEMO: Neutronic campaigns for T breeding optimization 用于EU DEMO的水冷铅和陶瓷增殖器（WLCB）繁殖毯（BB）：用于T育种优化的中子运动

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

Nuclear Materials and Energy

Pub Date : 2025-11-05 DOI: 10.1016/j.nme.2025.102022

Iole Palermo , Francisco A. Hernandez , Salvatore D’Amico , Jin Hun Park , Pavel Pereslavtsev , Guangming Zhou , Iñaki Zumalde

A novel Breeding Blanket (BB) concept, the Water-cooled Lead and Ceramic Breeder (WLCB) BB, has been developed under the EUROfusion Programme to address key challenges identified during the Pre-Conceptwual Design (PCD) phase for the driver BB concepts: Water-Cooled Lithium Lead (WCLL) and Helium-Cooled Pebble Bed (HCPB). The WLCB design represents an alternative hybrid approach, combining advantageous features of both WCLL and HCPB to mitigate their respective limitations: (1) shielding inefficiencies, challenging neutron multiplier technology, and integration concerns in HCPB; (2) challenges with tritium extraction from PbLi in WCLL variants; and (3) the reliance on anti-permeation barriers. In lieu of beryllium, alternative neutron multipliers—particularly lead—have been explored.

This work focuses on the neutronic optimization of the WLCB concept, with an emphasis on tritium breeding performance. Extensive neutronic simulation campaigns were conducted to optimize key design parameters, including toroidal and radial blanket layouts, cooling plate dimensions and water content, neutron multiplier zoning and materials (Pb, Be₁₂Ti, Zr₅Pb₃, C, ZrH₂), ⁶Li enrichment, ceramic breeder material, ceramic packing factor, and First Wall (FW) design to achieve the best results in terms of Tritium Breeding Ratio (TBR).

The resulting preliminary design − based on the best balance between neutronic performances and viability considerations, among others − achieves the EU DEMO TBR target of 1.15, representing a promising candidate for further development and integration into future design iterations.

一种新型的增殖毯（BB）概念，水冷铅和陶瓷增殖器（WLCB） BB，已经在欧洲聚变计划下开发出来，以解决在预概念设计（PCD）阶段确定的驱动BB概念的关键挑战：水冷锂铅（WCLL）和氦冷卵石床（HCPB）。WLCB的设计代表了一种替代的混合方法，结合了WCLL和HCPB的优点，以减轻各自的局限性：(1)屏蔽效率低下，具有挑战性的中子倍增器技术，以及HCPB中的集成问题；(2)从WCLL突变体PbLi中提取氚的挑战；(3)对防渗透屏障的依赖。除了铍之外，人们还探索了其他中子倍增器，尤其是铅。本工作着重于WLCB概念的中子优化，重点是氚的增殖性能。为优化关键设计参数，包括环形和径向包层布局、冷却板尺寸和含水率、中子倍增器分区和材料（Pb、Be12Ti、Zr5Pb3、C、ZrH2）、6Li富集、陶瓷增殖材料、陶瓷填充因子和第一壁（FW）设计，开展了大量中子模拟活动，以获得氚增殖比（TBR）方面的最佳结果。最终的初步设计——基于中子性能和可行性考虑之间的最佳平衡，以及其他因素——达到了欧盟DEMO的TBR目标1.15，代表了进一步开发和集成到未来设计迭代中的有希望的候选者。

{"title":"Water-cooled lead and ceramic breeder (WLCB) breeding blanket (BB) for the EU DEMO: Neutronic campaigns for T breeding optimization","authors":"Iole Palermo , Francisco A. Hernandez , Salvatore D’Amico , Jin Hun Park , Pavel Pereslavtsev , Guangming Zhou , Iñaki Zumalde","doi":"10.1016/j.nme.2025.102022","DOIUrl":"10.1016/j.nme.2025.102022","url":null,"abstract":"<div><div>A novel Breeding Blanket (BB) concept, the Water-cooled Lead and Ceramic Breeder (WLCB) BB, has been developed under the EUROfusion Programme to address key challenges identified during the Pre-Conceptwual Design (PCD) phase for the driver BB concepts: Water-Cooled Lithium Lead (WCLL) and Helium-Cooled Pebble Bed (HCPB). The WLCB design represents an alternative hybrid approach, combining advantageous features of both WCLL and HCPB to mitigate their respective limitations: (1) shielding inefficiencies, challenging neutron multiplier technology, and integration concerns in HCPB; (2) challenges with tritium extraction from PbLi in WCLL variants; and (3) the reliance on anti-permeation barriers. In lieu of beryllium, alternative neutron multipliers—particularly lead—have been explored.</div><div>This work focuses on the neutronic optimization of the WLCB concept, with an emphasis on tritium breeding performance. Extensive neutronic simulation campaigns were conducted to optimize key design parameters, including toroidal and radial blanket layouts, cooling plate dimensions and water content, neutron multiplier zoning and materials (Pb, Be<sub>12</sub>Ti, Zr<sub>5</sub>Pb<sub>3</sub>, C, ZrH<sub>2</sub>), <sup>6</sup>Li enrichment, ceramic breeder material, ceramic packing factor, and First Wall (FW) design to achieve the best results in terms of Tritium Breeding Ratio (TBR).</div><div>The resulting preliminary design − based on the best balance between neutronic performances and viability considerations, among others − achieves the EU DEMO TBR target of 1.15, representing a promising candidate for further development and integration into future design iterations.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"45 ","pages":"Article 102022"},"PeriodicalIF":2.7,"publicationDate":"2025-11-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145578614","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}

引用次数: 0