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Article(id=1320423083924636326, tenantId=1146029695717560320, journalId=1319355204128309394, issueId=1320423063842304050, articleNumber=null, orderNo=null, doi=10.3969/j.issn.1001-4837.2026.03.004, pmid=null, cstr=null, oa=null, hot=null, price=null, onlineType=0, articleFormat=0, articleType=null, articleTypeStr=null, receivedDate=1750780800000, receivedDateStr=2025-06-25, revisedDate=1762531200000, revisedDateStr=2025-11-08, acceptedDate=null, acceptedDateStr=null, onlineDate=1792742995481, onlineDateStr=2026-10-23, pubDate=1774800000000, pubDateStr=2026-03-30, doiRegisterDate=null, doiRegisterDateStr=null, onlineIssueDate=1792742995481, onlineIssueDateStr=2026-10-23, onlineJustAcceptDate=null, onlineJustAcceptDateStr=null, onlineFirstDate=null, onlineFirstDateStr=null, sourceXml=null, magXml=null, createTime=1792742995481, creator=13701087609, updateTime=1792742995481, updator=13701087609, issue=Issue{id=1320423063842304050, tenantId=1146029695717560320, journalId=1319355204128309394, year='2026', volume='43', issue='3', pageStart='1', pageEnd='86', issueExtLink='null', onlineDate='null', pubDate='1774800000000', pubDateStr='2026-03-30', beforeIssueId=null, nextIssueId=null, price=null, status=1, issueComplete=1, articleOrder=1, issueType=1, specialIssue=null, createTime=1792742990694, creator='13701087609', updateTime=1792745289363, updator='13701087609', preIssue=null, nextIssue=null, articleTotal=null, ext={EN=IssueExt(id=1320432705251165019, tenantId=1146029695717560320, journalId=1319355204128309394, issueId=1320423063842304050, language=EN, specialIssueTitle=, coverIllustrator=null, specialIssueEditor=, specialIssueAbout=), CN=IssueExt(id=1320432705251165020, tenantId=1146029695717560320, journalId=1319355204128309394, issueId=1320423063842304050, language=CN, specialIssueTitle=, coverIllustrator=null, specialIssueEditor=, specialIssueAbout=)}, issueFiles=null, downloadFileDto=null}, startPage=42, endPage=50, ext={EN=ArticleExt(id=1320423084159517351, articleId=1320423083924636326, tenantId=1146029695717560320, journalId=1319355204128309394, language=EN, title=Study on the high‑temperature irradiation behavior of austenitic stainless steel 316H for pressure vessels, columnId=1320422992476226524, journalTitle=Pressure Vessel Technology, columnName=Experimental Researches, runingTitle=null, highlight=null, articleAbstract=

To address the lack of high‑temperature irradiation performance data for the austenitic stainless steel 316H，the material used for the reactor pressure vessel of a fourth‑generation high‑temperature gas‑cooled microreactor，at the design operating temperature of 550 ℃，a research plan involving irradiation damage tests and aging tests was proposed.The SRIM software was used to simulate the trajectories and irradiation damage defects of energetic ions （He2+） incident on the material.Combined with high‑temperature irradiation tests under different cumulative irradiation doses （0.005，0.05，0.5 dpa） and an irradiation aging test at 650 ℃ for 100 h，the evolution laws of the material's high‑temperature damage behavior and aging behavior with microstructure were established.The results show that at the design temperature of 550 ℃，316H exhibits good irradiation resistance.The irradiation damage layer is approximately several micrometers thick，and the damage mainly manifests as microvoids and dislocations.After aging at 650 ℃ for 100 h，carbides precipitate along grain boundaries，with C and Mo elements enriching at the grain boundaries and Cr element depleting near the grain boundaries，presenting a certain risk of failure.This study provides a reference for subsequent engineering implementation of in‑reactor irradiation tests.

, authors=Zhe LI, Yuerong WANG, Shaoyang JIAO, authorsList=Zhe LI, Yuerong WANG, Shaoyang JIAO, authorCompany=null, correspAuthors=null, authorNote=null, correspAuthorsNote=null, copyrightStatement=null, copyrightOwner=null, extLink=null, articleAbsUrl=null, sourceXml=null, magXml=null, pdfUrl=null, pdf=null, pdfFileSize=null, pdfExtLink=null, richHtmlUrl=null, mobilePdfUrl=null, reviewReport=null, pdfFirstPage=null, abstractGraph=null, abstractGraphContent=null, abstractVideo=null, citation=null, cebUrl=null, magXmlContent=null, mapNumber=null, fund=null), CN=ArticleExt(id=1320423088118940350, articleId=1320423083924636326, tenantId=1146029695717560320, journalId=1319355204128309394, language=CN, title=压力容器用奥氏体不锈钢316H的高温辐照行为研究, columnId=1320422992698524638, journalTitle=压力容器, columnName=试验研究, runingTitle=null, highlight=null, articleAbstract=

针对某四代高温气冷微堆堆芯压力容器材料奥氏体不锈钢316H在设计运行温度550 ℃下缺乏高温辐照性能数据的问题，提出开展辐照损伤试验与老化试验的研究方案。通过SRIM软件模拟载能离子（He2+）入射材料后的运动轨迹与辐照损伤缺陷，结合不同累积辐照剂量下（0.005，0.05，0.5 dpa）的高温辐照试验及650 ℃、100 h的辐照老化试验，建立材料高温损伤行为及老化行为与微观组织的演化规律。结果表明：在设计温度550 ℃下，316H抗辐照性能良好，辐照损伤层约为几微米，损伤主要表现为微空洞和位错；在650 ℃下经100 h老化后，晶间析出碳化物，C和Mo元素在晶界处富集，Cr元素在晶界附近贫化，存在一定失效风险。研究为后续工程实施堆内辐照试验提供了参考。

, authors=李哲, 王跃蓉, 焦少阳, authorsList=李哲, 王跃蓉, 焦少阳, authorCompany=null, correspAuthors=null, authorNote=

李哲（1988），女，副高级工程师，主要从事核工程材料、焊接与无损检测设计方向的研究工作，通信地址：100840北京市海淀区首体南路9号中核工程办公楼，E-mail：。

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李哲（1988），女，副高级工程师，主要从事核工程材料、焊接与无损检测设计方向的研究工作，通信地址：100840北京市海淀区首体南路9号中核工程办公楼，E-mail：。

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李哲（1988），女，副高级工程师，主要从事核工程材料、焊接与无损检测设计方向的研究工作，通信地址：100840北京市海淀区首体南路9号中核工程办公楼，E-mail：。

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High‑temperature tensile properties of austeniticstainless steel 316H

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项目温度/℃
500550600650

屈服强度

/MPa

实测值137139141131
141144140129
146133135126
标准值≥118≥116≥114≥110

抗拉强度

/MPa

实测值454447422372
455442420368
459438378369
标准值≥420≥393≥357≥275
), ArticleFig(id=1320423092984333039, tenantId=1146029695717560320, journalId=1319355204128309394, articleId=1320423083924636326, language=CN, label=表1, caption=

奥氏体不锈钢316H的高温拉伸性能

, figureFileSmall=null, figureFileBig=null, tableContent=
项目温度/℃
500550600650

屈服强度

/MPa

实测值137139141131
141144140129
146133135126
标准值≥118≥116≥114≥110

抗拉强度

/MPa

实测值454447422372
455442420368
459438378369
标准值≥420≥393≥357≥275
), ArticleFig(id=1320423093072413424, tenantId=1146029695717560320, journalId=1319355204128309394, articleId=1320423083924636326, language=EN, label=Tab.2, caption=

Chemical composition of austenitic stainless steel 316H

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元素FeCrNiMo
原子百分比/%65.719.013.31.90
质量分数/%65.217.613.93.24
), ArticleFig(id=1320423093143716593, tenantId=1146029695717560320, journalId=1319355204128309394, articleId=1320423083924636326, language=CN, label=表2, caption=

奥氏体不锈钢316H的化学成分

, figureFileSmall=null, figureFileBig=null, tableContent=
元素FeCrNiMo
原子百分比/%65.719.013.31.90
质量分数/%65.217.613.93.24
), ArticleFig(id=1320423093202436850, tenantId=1146029695717560320, journalId=1319355204128309394, articleId=1320423083924636326, language=EN, label=Tab.3, caption=

Relationship between He2+ and fast neutron injectionamounts under equivalent cascade damage conditions

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级联损伤/dpa0.0050.050.5
快中子累计注量/（n·cm-2）1.0×10192.0×10192.0×1021
He2+注量/（ions·cm-2）1.4×10141.4×10151.4×1016
), ArticleFig(id=1320423093286322931, tenantId=1146029695717560320, journalId=1319355204128309394, articleId=1320423083924636326, language=CN, label=表3, caption=

同等级联损伤条件下He2+和快中子注量的关系

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级联损伤/dpa0.0050.050.5
快中子累计注量/（n·cm-2）1.0×10192.0×10192.0×1021
He2+注量/（ions·cm-2）1.4×10141.4×10151.4×1016
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压力容器用奥氏体不锈钢316H的高温辐照行为研究
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李哲 , 王跃蓉 , 焦少阳
压力容器 | 试验研究 2026,43(3): 42-50
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压力容器 |试验研究 2026 , 43 (3) : 42 -50
压力容器用奥氏体不锈钢316H的高温辐照行为研究
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李哲 , 王跃蓉, 焦少阳
作者信息
  • 中国核电工程有限公司，北京 100840
作者简介:

李哲（1988），女，副高级工程师，主要从事核工程材料、焊接与无损检测设计方向的研究工作，通信地址：100840北京市海淀区首体南路9号中核工程办公楼，E-mail：。

Study on the high‑temperature irradiation behavior of austenitic stainless steel 316H for pressure vessels
Zhe LI , Yuerong WANG, Shaoyang JIAO
Affiliations
  • China Nuclear Power Engineering Co.，Ltd.，Beijing 100840，China
出版时间: 2026-03-30 doi: 10.3969/j.issn.1001-4837.2026.03.004
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摘要
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针对某四代高温气冷微堆堆芯压力容器材料奥氏体不锈钢316H在设计运行温度550 ℃下缺乏高温辐照性能数据的问题，提出开展辐照损伤试验与老化试验的研究方案。通过SRIM软件模拟载能离子（He2+）入射材料后的运动轨迹与辐照损伤缺陷，结合不同累积辐照剂量下（0.005，0.05，0.5 dpa）的高温辐照试验及650 ℃、100 h的辐照老化试验，建立材料高温损伤行为及老化行为与微观组织的演化规律。结果表明：在设计温度550 ℃下，316H抗辐照性能良好，辐照损伤层约为几微米，损伤主要表现为微空洞和位错；在650 ℃下经100 h老化后，晶间析出碳化物，C和Mo元素在晶界处富集，Cr元素在晶界附近贫化，存在一定失效风险。研究为后续工程实施堆内辐照试验提供了参考。

高温反应堆  /  316H  /  高温辐照损伤

To address the lack of high‑temperature irradiation performance data for the austenitic stainless steel 316H，the material used for the reactor pressure vessel of a fourth‑generation high‑temperature gas‑cooled microreactor，at the design operating temperature of 550 ℃，a research plan involving irradiation damage tests and aging tests was proposed.The SRIM software was used to simulate the trajectories and irradiation damage defects of energetic ions （He2+） incident on the material.Combined with high‑temperature irradiation tests under different cumulative irradiation doses （0.005，0.05，0.5 dpa） and an irradiation aging test at 650 ℃ for 100 h，the evolution laws of the material's high‑temperature damage behavior and aging behavior with microstructure were established.The results show that at the design temperature of 550 ℃，316H exhibits good irradiation resistance.The irradiation damage layer is approximately several micrometers thick，and the damage mainly manifests as microvoids and dislocations.After aging at 650 ℃ for 100 h，carbides precipitate along grain boundaries，with C and Mo elements enriching at the grain boundaries and Cr element depleting near the grain boundaries，presenting a certain risk of failure.This study provides a reference for subsequent engineering implementation of in‑reactor irradiation tests.

high‑temperature reactor  /  316H  /  high‑temperature irradiation damage
李哲, 王跃蓉, 焦少阳. 压力容器用奥氏体不锈钢316H的高温辐照行为研究. 压力容器, 2026 , 43 (3) : 42 -50 . DOI: 10.3969/j.issn.1001-4837.2026.03.004
Zhe LI, Yuerong WANG, Shaoyang JIAO. Study on the high‑temperature irradiation behavior of austenitic stainless steel 316H for pressure vessels[J]. Pressure Vessel Technology, 2026 , 43 (3) : 42 -50 . DOI: 10.3969/j.issn.1001-4837.2026.03.004
正文
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0　引言
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随着“双碳”目标的提出，我国核电行业进入了新一轮的快速发展期。在第四代核电技术领域，山东石岛湾高温气冷堆示范工程已于2021年12月实现并网发电［1］；福建霞浦钠冷快堆示范工程1号机组已建成发电，2号机组正在建设中；钍基熔盐实验堆也已建成并装料，正在开展相关的试验验证工作。
反应堆压力容器（reactor pressure vessel，RPV）是核电站反应堆冷却剂系统压力边界的重要组成部分，既承受反应堆冷却剂压力，也作为反应堆放射性产物的第二道屏障。除承受各种机械载荷外，还承受高温、高压及强辐照的作用，其完整性直接影响反应堆的安全与寿命。目前，国内核电机组三代压力容器用钢一般选用SA-508 Gr.3 Cl.1或与之等效的16MND5材料，行业内针对该类材料的性能［2］、机械性能与韧性［3］、焊接性能［4-5］、辐照性能［6-7］等已有相当深入的研究和验证。
四代核电堆型由于运行温度远高于三代压水堆，主回路系统设备材料不可避免地在高温服役环境中承受长期高通量中子辐照，以及嬗变反应产生的He2+的作用。这些作用必然导致合金微观结构发生一定变化，主要表现为移位损伤［8-9］。在现有的四代堆中，钍基熔盐堆具有压力小、温度高的特点，主要考虑高温熔盐腐蚀问题，选用自研国产哈氏合金GH3535（相当于ASME牌号UNS N10003）作为反应堆压力容器承压边界材料；钠冷快堆的反应堆压力边界需保证高温结构完整性，考虑高温钠腐蚀和快中子辐照的影响，选用自研国产奥氏体不锈钢316KD（相当于ASME 316H）作为反应堆压力容器承压边界材料。
在文中涉及的某四代高温气冷微堆研发项目中，反应堆压力容器的设计温度为550 ℃，运行壁温为510 ℃，运行压力约为3 MPa，设计寿命为20年，E>1 MeV时的快中子累计注量为1×1019 n/cm2。结合ASME BPVC-Ⅲ-5—2015 Rules for Construction of Nuclear Power Plant Components，Division 5：High Temperature Reactors中HBB部分对材料的相关要求，推荐304、304H、316、316H四个牌号的奥氏体不锈钢作为承压边界材料。鉴于H级别不锈钢的使用温度更高，因此优先考虑采用高温奥氏体不锈钢材料316H，该材料与钠冷快堆压力容器所用材料牌号类似。由于钠冷快堆的压力容器设计运行温度为440 ℃、压力为0.06 MPa，而对于其在550 ℃设计运行温度下，特别是承受高通量中子辐照（约1×1019 n/cm2，E>1 MeV）后的长期性能，如微观组织稳定性、力学性能退化、辐照促进析出与偏聚行为等，目前尚缺乏充分的试验数据支撑。
对316H开展了不同温度下的高温力学性能试验，结果如表1所示。在设计温度及极限工况温度下，316H的抗拉强度和屈服强度均满足设计及ASME BPVC-Ⅲ-5—2015要求，初步满足应用的力学强度要求。
从工程应用角度出发，四代堆型选用316H材料除了应满足设计温度下的机械性能、高温持久强度、蠕变强度和疲劳强度外，还需考虑其在设计温度下的高温老化、辐照损伤等行为特性。高温辐照损伤试验常因条件限制而难以在短时间内开展，因此利用计算机进行模拟辐照损伤研究的进展更为迅速。FRISONI等［10］、SHIMAKAWA等［11］分别利用不同的模型计算了堆用结构材料的辐照损伤和He2+产生速率。文献［12-15］通过模型预测或试验验证的方式对奥氏体不锈钢的辐照性能进行了研究。值得注意的是，黄鹤飞等［16］对四代熔盐堆用奥氏体不锈钢316SS进行了离子辐照损伤中的温度效应研究，研究表明室温下氙离子辐照使得316SS内部会形成大量纳米尺度的位错环，进而使材料产生辐照硬化。
因此，本文针对奥氏体不锈钢材料316H，开展服役工况下不同辐照剂量的性能试验，研究设计温度550 ℃下的模拟中子辐照特性，并开展650 ℃高温辐照老化试验，建立辐照损伤机理与微观组织的内在联系，探究高温老化行为与微观组织的演化规律，揭示潜在的微观失效机制，为该材料在后续工程上的应用提供一定的数据支持。
1　级联损伤计算
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采用SRIM 2008软件包，基于蒙特卡洛方法模拟载能离子（He2+）入射奥氏体不锈钢的运动轨迹，并对辐照损伤缺陷进行统计，计算不同入射深度下的级联损伤程度。
将奥氏体不锈钢316H的化学成分作为输入（见表2），材料密度为7.9 g/cm3。载能离子初始条件如下：540 keV He离子，垂直入射到316H块体内部，设定的金属层厚度为2 μm。对入射后不同深度的空位数密度进行统计，可得到累积辐照剂量随表面深度的分布信息，其中入射面和截面方向的离子轨迹如图1所示。
一般可以采用高能离子代替中子来模拟研究中子辐照对原子的离位损伤。由于He2+易于获得，同时质量与中子比较接近，选用He2+不仅能模拟中子产生的离位损伤，还能引入嬗变氢和氦效应，更全面地模拟实际中子辐照环境。因此，文中选用He2+作为载能离子模拟中子辐照损伤。向初始动能为540 keV的He2+中分别注入1.4×1014、1.4×1015和1.4×1016离子量的He2+，所产生的级联损伤分别为0.005，0.05，0.5 dpa，如图2所示。通常，在样品表面的浅表层（0~0.6 μm）范围内，辐照损伤变化较为缓慢；当深度大于0.6 μm后，样品的损伤程度迅速增加，并在1.0 μm左右达到最大值；随着深度的进一步增加，损伤强度快速衰减，在达到1.2 μm深度后衰减至零。
通常，辐照剂量的换算关系为1×1022 n/cm2相当于5.1 dpa。根据项目辐照剂量要求，压力容器辐照剂量为设计寿命20年，E>1 MeV的快中子累计注量约为1×1019 n/cm2。基于上述换算关系，累积辐照剂量为0.005 dpa。为便于比较，在同等级联损伤条件下，模拟载能粒子（He2+）与快中子注入量之间的换算关系如表3所示。
2　辐照损伤试验
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2.1　试样制备和试验条件
试样采用电火花线切割方法，从316H轧件上横向切割为15 mm×15 mm×2 mm的样品，对样品进行打磨和抛光，以去除残余应变层，然后采用银导电胶（液体）将样品粘贴在加热台上。待导电胶凝固后，将装有样品的载物台装入加热辐照舱内部，抽出辐照舱内部的空气至10-5 Pa量级，开启加热，设定加热目标温度为550±5 ℃。采用K型热电偶进行内部温度监测，待温度达到550 ℃后，稳定1 h再进行辐照。采用He2+作为载能粒子，粒子动能为540 keV，分别对样品进行辐照实验，按照表1设置3组不同的辐照剂量，分别对应0.005，0.05，0.5 dpa。
将辐照后的样品切割成2块2 mm×2 mm的样品，并将两个样品的辐照面彼此对粘，固化处理后获得对粘样品，如图3（a）所示。使用金刚石线将对粘样品切割为两个形状相似的样品，如图3（b）所示。将切割后的样品粘贴于带孔的钼环上，确保辐照面垂直于钼环所在的平面，如图3（c）所示。使用Gatan凹坑仪将样品表面减薄25~35 μm，使样品剩余厚度为15~25 μm。然后，将样品装入Gatan PIPS II 695聚焦型离子减薄仪中，采用的加速电压为5 kV，在倾角为9°的条件下减薄1 h，再将倾角调至7°减薄1 h。继而在倾角为5°条件下减薄，直至出孔，如图3（d）所示。最后，在倾角为4°条件下精修薄区，选用3.5 kV加速电压进一步扩大薄区，制备用于TEM观察的样品。
2.2　光学显微结果
对不同剂量辐照后的试样表面进行光学显微观察，如图4所示。可以看出，与空白试样对照组〔见图4（a）〕相比，试样在经过不同辐照剂量辐照后，表面光学形貌特征没有发生明显变化，表明损伤更为微观。因此，后续仅针对0.5 dpa辐照剂量下的样品表面进行进一步的SEM和TEM观察。
2.3　SEM形貌
采用SEM的背散射电子成像模式，分别观察0.5 dpa注入量前后试样的表面微观组织。图5示出试样的原始表面及不同条件辐照损伤后的表面形貌。图5（a）为原始样品的表面形貌。可以看出，晶粒尺寸约为30~70 μm，晶粒形貌大致呈等轴状。图5（b）为550 ℃下经0.5 dpa辐照损伤后的表面形貌，此时晶粒尺寸为20~70 μm，且大致呈等轴状，但未观察到明显损伤。图5（c）为试样在室温条件下经0.5 dpa辐照损伤后的表面形貌，相比于原始样品，辐照后样品表面无明显变化。
2.4　TEM观察
对减薄后的样品进行TEM观察。图6（a）示出奥氏体不锈钢316H经0.5 dpa辐照剂量辐照前后的TEM明场像，辐照截面浅表层可观察到微孔洞和位错环形貌。
由于辐照损伤导致衬度复杂，现有TEM图像已是所能获得的最佳状态。选中图6（a）中蓝色方框b区域进行放大，由图6（b）可以看出，辐照后样品内部形成微小孔洞，其尺寸为1~5 nm。同时，在［110］带轴双束条件下，可观察到Frank位错环，位错环的柏氏矢量如图6（c）所示。图6（d）示出Frank位错环的高分辨像，对其进行快速傅里叶变换（fast fourier transform，FFT）操作可获得图6（e），其中的周期性频域斑点与图6（b）（c）中的衍射花样一致。选取图中圆圈处的斑点（11¯1）和（1¯11¯），可对其他位置进行掩蔽。图6（f）示出掩蔽后经逆快速傅里叶变换（inverse fast fourier transform，IFFT）操作获得的图像，其中T字形标注为相应的位错缺陷，圈出的区域为位错环附近的畸变区。
2.5　辐照后的老化行为
对经0.005 dpa辐照损伤后的奥氏体316H不锈钢样品，在650 ℃下老化100 h。将预辐照老化后的样品用环氧树脂镶嵌，待其固化后对样品的截面进行打磨、抛光。为更好地反映辐照损伤层老化后的微观组织，在60%硝酸水溶液中的腐蚀时间约为5秒，略低于常规老化后试样金相观察的腐蚀时间。
图7示出辐照后老化的截面光学显微图像。可以看出，在样品浅表层1~1.5 μm范围内存在较大尺寸的颗粒状析出相，该范围与SRIM模拟预测的主要损伤区域深度范围（约0.6~1.2 μm）基本吻合（见图1）。
图8示出辐照试样经650 ℃/100 h老化后的SEM及EDS表征结果。
可以看出，试样经高温老化后，表面生成了厚度不足1 μm的氧化物。电解蚀刻后的SEM形貌观察表明，在浅表层范围内，碳化物析出相在部分晶界处选择性析出，这与奥氏体不锈钢的老化行为基本相似。图8（d）为不锈钢近表层的元素分布图，可以看出，镍元素在基体和晶界处的浓度相差不大，铁和铬元素在晶界部位出现明显贫化，碳元素在晶界部位呈现明显富集，钼元素仅在局部晶界出现富集现象。
图9示出近表面（5 μm以内）区域晶界及其邻近位置的元素分布。可以看出，在扫描线上，碳元素存在明显的局部富集，而铁、镍和锰等元素存在少量贫化，在远离晶界的位置，元素含量较为均匀，表明高温老化可造成晶界位置碳化物的析出。
图10（a）（b）为650 ℃/100 h老化后辐照样品截面靠近芯部位置的BSE高低倍微观组织图。可以看出，在晶界部位能够看到呈连续分布的析出相，相比基体内部呈现亮白色，表明析出相含有原子质量较大的元素。早期研究表明，在BSE成像模式下，亮白色的晶界主要为碳化物析出相。同时，部分晶粒内部含有孪晶界，且绝大部分孪晶界并不呈亮白色，如图10（b）（c）所示。因此，表明孪晶界处难以析出碳化物。图10（d）为截面近芯部区域的BSE观察及EDS元素分布的面扫结果。可以看出，在内部区域，C在晶界处较为富集，Mo元素在局部晶界较为富集，而Cr元素在晶界附近较为贫化。图11示出辐照试样芯部区域晶界附近的元素分布线扫结果，其结果基本与面扫结果一致。
辐照老化结果表明，晶界附近存在明显的C，Mo元素富集和Cr元素贫化，而Cr贫化是奥氏体不锈钢应力腐蚀失效和疲劳失效的重要原因之一，这种元素偏聚对材料长期高温服役安全性（如耐蚀性、韧性）构成潜在风险［17-19］。值得注意的是，部分晶粒内部的孪晶界处难以析出碳化物，这可能是由于孪晶界面能较低，同时其原子排列有序度高，不利于Cr元素的扩散和形核，这可为后续项目研发材料性能优化提供思路和参考方向。
3　结论
收起
文中通过辐照模拟计算和试样辐照老化试验，观察了奥氏体不锈钢316H辐照老化后的微观组织形貌变化。主要结论如下。
（1）奥氏体不锈钢316H在550 ℃、辐照剂量0.5 dpa的条件下辐照后，其显微组织未发生明显变化，证明该材料在高温条件下具备一定的抗中子辐照性能，辐照后的损伤形式主要为微空洞和位错环。
（2）经0.005 dpa辐照损伤后的奥氏体不锈钢316H样品，在650 ℃下老化100 h后，晶界处出现碳化物析出相，部分晶粒内部含有孪晶界。C在晶界处较为富集，Mo元素在局部晶界较为富集，而Cr元素在晶界附近较为贫化。
（3）奥氏体不锈钢316H在650 ℃高温老化后，由于晶界碳化物析出伴随显著的Cr元素贫化，存在晶间腐蚀和脆化的潜在风险。因此，在工程应用评估中，特别是针对接近或可能达到650 ℃的工况，建议实施堆内辐照监督计划，并开展更接近实际辐照环境的堆内中子辐照试验，以全面评估其长期性能退化行为。

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doi: 10.3969/j.issn.1001-4837.2026.03.004
  • 接收时间：2025-06-25
  • 首发时间：2026-10-23
  • 出版时间：2026-03-30
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  • 收稿日期：2025-06-25
  • 修回日期：2025-11-08
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2种不同金属材料的力学参数
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鹅膏菌科Amanitaceae 2 11 5.26 鹅膏菌属 Amanita 10 4.78
小菇科 Mycenaceae 2 12 5.74 丝盖伞属 Inocybe 5 2.39
多孔菌科 Polyporaceae 8 14 6.70 蜡蘑属 Laccaria 5 2.39
红菇科 Russulaceae 3 23 11.00 小皮伞属 Marasmius 6 2.87
小菇属 Mycena 11 5.26
光柄菇属 Pluteus 5 2.39
红菇属 Russula 17 8.13
栓菌属 Trametes 5 2.39
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