![Fig. 10. Summary of US LWR burn-up and capacity factors [78], [79], [80]](https://ars.els-cdn.com/content/image/1-s2.0-S1359645412007987-gr10.sml)
![Fig. 16. Comparison of the volumetric void swelling behavior of Type 304L [116], 316…](https://ars.els-cdn.com/content/image/1-s2.0-S1359645412007987-gr16.sml)
Materials challenges in nuclear energy
核能中的材料挑战
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Abstract 摘要
Nuclear power currently provides about 13% of electrical power worldwide, and has emerged as a reliable baseload source of electricity. A number of materials challenges must be successfully resolved for nuclear energy to continue to make further improvements in reliability, safety and economics. The operating environment for materials in current and proposed future nuclear energy systems is summarized, along with a description of materials used for the main operating components. Materials challenges associated with power uprates and extensions of the operating lifetimes of reactors are described. The three major materials challenges for the current and next generation of water-cooled fission reactors are centered on two structural materials aging degradation issues (corrosion and stress corrosion cracking of structural materials and neutron-induced embrittlement of reactor pressure vessels), along with improved fuel system reliability and accident tolerance issues. The major corrosion and stress corrosion cracking degradation mechanisms for light-water reactors are reviewed. The materials degradation issues for the Zr alloy-clad UO2 fuel system currently utilized in the majority of commercial nuclear power plants are discussed for normal and off-normal operating conditions. Looking to proposed future (Generation IV) fission and fusion energy systems, there are five key bulk radiation degradation effects (low temperature radiation hardening and embrittlement; radiation-induced and -modified solute segregation and phase stability; irradiation creep; void swelling; and high-temperature helium embrittlement) and a multitude of corrosion and stress corrosion cracking effects (including irradiation-assisted phenomena) that can have a major impact on the performance of structural materials.
核能目前为全球电力供应提供了约 13%，已成为可靠的基荷电力来源。为了使核能继续在可靠性、安全性和经济性方面取得进一步改进，必须成功解决一系列材料挑战。本文总结了当前及未来核能系统中材料的工作环境，并描述了用于主要运行部件的材料。文中还描述了与反应堆功率提升和运行寿命延长相关的材料挑战。目前及下一代水冷裂变反应堆的三大材料挑战主要集中在对两种结构材料的两个老化降解问题（结构材料的腐蚀和应力腐蚀开裂，以及反应堆压力容器的中子辐照脆化），以及改进燃料系统可靠性和事故耐受性问题。文中回顾了轻水反应堆的主要腐蚀和应力腐蚀开裂降解机制。 目前商用核电站普遍使用的 Zr 合金包壳 UO 燃料系统在正常和异常运行条件下的材料退化问题进行了讨论。展望未来的（第四代）裂变和聚变能源系统，存在五种主要的体辐射退化效应（低温辐射硬化和脆化；辐射诱导和改性溶质偏析及相稳定性；辐照蠕变；空位肿胀；高温氦脆化）以及多种腐蚀和应力腐蚀开裂效应（包括辐照辅助现象），这些都会对结构材料的性能产生重大影响。
核能目前为全球电力供应提供了约 13%，已成为可靠的基荷电力来源。为了使核能继续在可靠性、安全性和经济性方面取得进一步改进，必须成功解决一系列材料挑战。本文总结了当前及未来核能系统中材料的工作环境，并描述了用于主要运行部件的材料。文中还描述了与反应堆功率提升和运行寿命延长相关的材料挑战。目前及下一代水冷裂变反应堆的三大材料挑战主要集中在对两种结构材料的两个老化降解问题（结构材料的腐蚀和应力腐蚀开裂，以及反应堆压力容器的中子辐照脆化），以及改进燃料系统可靠性和事故耐受性问题。文中回顾了轻水反应堆的主要腐蚀和应力腐蚀开裂降解机制。 目前商用核电站普遍使用的 Zr 合金包壳 UO 燃料系统在正常和异常运行条件下的材料退化问题进行了讨论。展望未来的（第四代）裂变和聚变能源系统，存在五种主要的体辐射退化效应（低温辐射硬化和脆化；辐射诱导和改性溶质偏析及相稳定性；辐照蠕变；空位肿胀；高温氦脆化）以及多种腐蚀和应力腐蚀开裂效应（包括辐照辅助现象），这些都会对结构材料的性能产生重大影响。
Keywords 关键词
Nuclear materials
Radiation effects
Stress corrosion cracking
Structural alloys (steels and nickel base)
Nuclear fuels
核材料 辐射效应 应力腐蚀开裂 结构合金（钢和镍基） 核燃料
1. Introduction 1. 引言
Access to reliable, sustainable and affordable energy is viewed as crucial to worldwide economic prosperity and stability [1], [2]. Nuclear fission energy has emerged over the past 40 years to become a reliable baseload source of clean and economical electrical energy. As of 2011, there were 435 nuclear reactors in operation worldwide, producing 370 GWe of electricity [3]. Another 108 units or 108 GWe are forthcoming (under construction or on order), for a total of 543 units and 478 GWe of electrical capacity. The largest producer of power from nuclear energy is the USA, with 104 commercial reactors licensed to operate at 65 sites, producing a total of 103 GWe of electricity. These provided just under 20% of the nation’s total electric energy generation and more than 30% of worldwide nuclear generating capacity. Worldwide, nuclear energy provides about 13% of the electrical demand [1]. Given that nuclear power has very low carbon emission [2] and that energy generation currently accounts for 66% of worldwide greenhouse gas emissions [4], nuclear energy is considered an important resource in managing atmospheric greenhouse gases and associated climate change [1].
获得可靠、可持续且负担得起的能源被视为对全球经济发展和稳定至关重要[1],[2]。核裂变能源在过去 40 年间已成为一种可靠的基础电力来源，提供清洁且经济的电力。截至 2011 年，全球有 435 座核反应堆在运行，发电量为 370 吉瓦 e 。另外还有 108 台机组或 108 吉瓦 e 正在建设或订购中，总共有 543 台机组和 478 吉瓦 e 的电力装机容量。美国是核能发电量最大的国家，有 104 座商业核反应堆获准在 65 个地点运行，总共发电量为 103 吉瓦 e 。这些核反应堆提供了全国近 20%的电力总发电量，以及全球超过 30%的核发电能力。全球范围内，核能满足了约 13%的电力需求[1]。鉴于核能的碳排放量非常低[2]，而目前能源生产占全球温室气体排放的 66%[4]，核能被认为是管理大气中温室气体及相关气候变化的重要资源[1]。
获得可靠、可持续且负担得起的能源被视为对全球经济发展和稳定至关重要[1],[2]。核裂变能源在过去 40 年间已成为一种可靠的基础电力来源，提供清洁且经济的电力。截至 2011 年，全球有 435 座核反应堆在运行，发电量为 370 吉瓦 e 。另外还有 108 台机组或 108 吉瓦 e 正在建设或订购中，总共有 543 台机组和 478 吉瓦 e 的电力装机容量。美国是核能发电量最大的国家，有 104 座商业核反应堆获准在 65 个地点运行，总共发电量为 103 吉瓦 e 。这些核反应堆提供了全国近 20%的电力总发电量，以及全球超过 30%的核发电能力。全球范围内，核能满足了约 13%的电力需求[1]。鉴于核能的碳排放量非常低[2]，而目前能源生产占全球温室气体排放的 66%[4]，核能被认为是管理大气中温室气体及相关气候变化的重要资源[1]。
The core of a nuclear reactor presents an exceptionally harsh environment for materials due to the combination of high temperature, high stresses, a chemically aggressive coolant and intense radiation fluxes. Many of the features that make reactors attractive from a physics perspective (e.g. high specific power, self-sustaining reaction) exert high operational burdens on structural materials. For example, the recoverable energy from each 235U fission reaction is ∼200 MeV, which is about eight orders of magnitude per atom higher than typical chemical reactions. As a result, typical power densities in commercial nuclear reactor cores are ∼50–75 MWth m−3, which is nearly two orders of magnitude higher than the average power density in the boiler furnace of a large-scale coal power plant. This intense production of heat is accompanied by the generation of energetic neutrons (which serve to sustain the fission reaction) and gamma radiation, which can degrade materials by displacement damage and radiolysis processes, respectively. Recent activities to extend the operating lifetime of current water reactors, to develop advanced fission reactor concepts with greater functionality and capability, and the coming emergence of fusion energy represent even greater demands on materials [5], [6], [7], [8].
核反应堆的核心环境对材料来说极为苛刻，这是由于高温、高应力、化学性质具有强腐蚀性的冷却剂以及强烈的辐射通量共同作用的结果。从物理学角度来看，许多使反应堆具有吸引力的特性（例如高比功率、自持反应）对结构材料施加了极高的运行负担。例如，每个铀-235 裂变反应可回收的能量约为 200 兆电子伏特，这比典型的化学反应每个原子高出八个数量级。因此，商业核反应堆核心的典型功率密度约为 50-75 兆瓦/立方米，这比大型煤电厂锅炉炉膛的平均功率密度高出近两个数量级。这种剧烈的热量产生伴随着高能中子（这些中子用于维持裂变反应）和伽马射线的产生，分别通过位移损伤和辐射解离过程使材料退化。 近期活动旨在延长现有水堆的运行寿命，开发功能更强、能力更高的先进裂变反应堆概念，以及聚变能源的即将出现，都对材料提出了更高的要求[5]，[6]，[7]，[8]。
核反应堆的核心环境对材料来说极为苛刻，这是由于高温、高应力、化学性质具有强腐蚀性的冷却剂以及强烈的辐射通量共同作用的结果。从物理学角度来看，许多使反应堆具有吸引力的特性（例如高比功率、自持反应）对结构材料施加了极高的运行负担。例如，每个铀-235 裂变反应可回收的能量约为 200 兆电子伏特，这比典型的化学反应每个原子高出八个数量级。因此，商业核反应堆核心的典型功率密度约为 50-75 兆瓦/立方米，这比大型煤电厂锅炉炉膛的平均功率密度高出近两个数量级。这种剧烈的热量产生伴随着高能中子（这些中子用于维持裂变反应）和伽马射线的产生，分别通过位移损伤和辐射解离过程使材料退化。 近期活动旨在延长现有水堆的运行寿命，开发功能更强、能力更高的先进裂变反应堆概念，以及聚变能源的即将出现，都对材料提出了更高的要求[5]，[6]，[7]，[8]。
1.1. Types of nuclear fission reactors
1.1. 核裂变反应堆类型
The predominant reactor design worldwide is the pressurized water reactor (PWR), accounting for two-thirds of the installed capacity, followed by boiling water reactors (BWRs) at 21% and heavy-water reactors at 14% of installed capacity, respectively (Table 1) [3]. All of these water-cooled reactors use ceramic fuel pellets consisting of UO2 or other fissile actinide oxides to generate heat. The ceramic pellets are stacked inside of long Zr alloy tubes (fuel cladding) that transfer the nuclear heat to flowing water coolant and serve as the primary barrier containing the volatile radioactive fission byproducts. The remaining 5% of installed nuclear energy comes from gas-cooled reactors, graphite-moderated reactors and liquid metal cooled reactors (Table 1).
全球主要的反应堆设计是压水堆（PWR），占装机容量的三分之二，其次是沸水堆（BWR）占 21%，重水堆分别占装机容量的 14%（表 1）[3]。所有这些水冷反应堆都使用由 UO₂或其他可裂变锕系氧化物制成的陶瓷燃料芯块来产生热量。陶瓷芯块堆叠在长锆合金管（燃料包壳）内，这些管将核热传递给流动的水冷却剂，并作为主要屏障容纳挥发性放射性裂变产物。其余 5%的核能装机容量来自气冷反应堆、石墨慢化反应堆和液态金属冷却反应堆（表 1）。
全球主要的反应堆设计是压水堆（PWR），占装机容量的三分之二，其次是沸水堆（BWR）占 21%，重水堆分别占装机容量的 14%（表 1）[3]。所有这些水冷反应堆都使用由 UO₂或其他可裂变锕系氧化物制成的陶瓷燃料芯块来产生热量。陶瓷芯块堆叠在长锆合金管（燃料包壳）内，这些管将核热传递给流动的水冷却剂，并作为主要屏障容纳挥发性放射性裂变产物。其余 5%的核能装机容量来自气冷反应堆、石墨慢化反应堆和液态金属冷却反应堆（表 1）。
Table 1. Power reactors by type, worldwide [3].
表 1. 全球按类型划分的发电反应堆[3]。
| Reactor type 反应堆类型 | # Units # 单位 | Net MWe 净兆瓦 e | # Units # 单位 | Net MWe 净兆瓦 e | # Units # 单位 | Net MWe 净兆瓦 e |
|---|---|---|---|---|---|---|
| Empty Cell | (in operation) （运行中） | Empty Cell | (forthcoming) | Empty Cell | (total) | Empty Cell |
| Pressurized light-water reactors (PWR) 压水堆 | 267 | 246555.1 | 89 | 93,014 | 356 | 339569.1 |
| Boiling light-water reactors (BWR) 沸水堆 | 84 | 78320.6 | 6 | 8056 | 90 | 86376.6 |
| Gas-cooled reactors, all models 气体冷却反应堆，所有型号 | 17 | 8732.0 | 1 | 200 | 18 | 8932.0 |
| Heavy-water reactors, all models 重水反应堆，所有型号 | 51 | 25610.0 | 8 | 5112 | 59 | 30722.0 |
| Graphite-moderated reactors, all models 石墨慢化反应堆，所有型号 | 15 | 10219.0 | 0 | 0 | 15 | 10219.0 |
| Liquid-metal-cooled reactors, all models 液态金属冷却反应堆，所有型号 | 1 | 560.0 | 4 | 1016 | 5 | 2076.0 |
| Totals 总计 | 435 | 369996.7 | 108 | 107,896 | 543 | 477894.7 |
The vast majority of the reactors listed in Table 1 are classified as Generation II reactors [9], which were designed in the 1960s and predominantly achieved initial commercial operation from the 1970s through the 1990s. These reactors are distinguished from Generation I designs (1950s-60s), which were early commercial prototype and demonstration reactors, and Generation III reactors, designed in the 1990s to incorporate significant advances in safety and economics [9]. Generation III reactor construction for the past decade has been centered in Asia, with a few units recently built in Europe. The current generation of light-water reactors (LWRs), Generation III+, include still further advancement in economics and safety, such as passive heat removal systems. There are a total of 108 Generation III and Generation III+ reactors on order or under construction around the world, and of those, 89 are PWRs.
表 1 中列出的绝大多数反应堆被归类为第二代反应堆[9]，这些反应堆设计于 20 世纪 60 年代，主要在 20 世纪 70 年代至 90 年代实现了初步商业运行。这些反应堆与第一代设计（20 世纪 50-60 年代）区分开来，第一代设计是早期的商业原型和示范反应堆，以及设计于 20 世纪 90 年代的第三代反应堆，后者旨在融入显著的安全和经济进步[9]。过去十年中，第三代反应堆的建设主要集中在亚洲，欧洲最近也建造了一些。当前一代轻水反应堆（LWRs），即第三代+反应堆，在经济和安全方面仍有进一步发展，例如采用被动冷却系统。全球共有 108 座第三代和第三代+反应堆正在订购或建设中，其中 89 座是压水堆。
表 1 中列出的绝大多数反应堆被归类为第二代反应堆[9]，这些反应堆设计于 20 世纪 60 年代，主要在 20 世纪 70 年代至 90 年代实现了初步商业运行。这些反应堆与第一代设计（20 世纪 50-60 年代）区分开来，第一代设计是早期的商业原型和示范反应堆，以及设计于 20 世纪 90 年代的第三代反应堆，后者旨在融入显著的安全和经济进步[9]。过去十年中，第三代反应堆的建设主要集中在亚洲，欧洲最近也建造了一些。当前一代轻水反应堆（LWRs），即第三代+反应堆，在经济和安全方面仍有进一步发展，例如采用被动冷却系统。全球共有 108 座第三代和第三代+反应堆正在订购或建设中，其中 89 座是压水堆。
Given the high representation of PWRs and BWRs in the world’s fleet, materials issues in these two types of reactors are of greatest interest. And of the many materials in a reactor, those that experience the most extreme conditions (stress, corrosion, and radiation) are most important for maintaining plant safety and reliability. Fig. 1 shows a schematic of the major components in the primary and secondary circuits of a PWR [10]. Pressurized water (∼15.5 MPa) in the primary circuit enters the reactor core at ∼275 °C, picks up heat from the reactor core with a core exit temperature of ∼325 °C, and transfers the heat across the U-tubes in the steam generator to water at a lower pressure. This water turns to steam that powers the turbine, and is condensed and recirculated. Fig. 1 also lists the alloys used throughout the primary and secondary circuits, all of which are in contact with high-temperature water and are subject to significant mechanical stress. Alloys inside (and including) the reactor vessel are also subject to varying levels of radiation, which produces displacement damage and radiolytic decomposition of the coolant water. Major pressure boundary components (reactor pressure vessel, pressurizer, steam generator, steam lines, turbine and condenser) are made of either low carbon or low alloy steel. Austenitic stainless steels (Types 304, 304L, 316, 316L, 321, 347) dominate the core structural materials, as well as serving for cladding (308SS and 309SS) on the inside surface of the reactor pressure vessel and pressurizer. Higher strength components such as springs and fasteners are made of nickel-base alloys. Vessel penetrations and steam generator tubes are made of nickel-base alloy 690 (previously alloy 600, which was found to provide insufficient resistance to stress corrosion cracking). Condenser tubes are generally made of titanium or stainless steel. The selection of nickel-base alloys and austenitic stainless steels for core internals and the steam generator tubes is driven by the need for good aqueous corrosion resistance at high temperatures. These alloys have low corrosion rates due to the formation of chromium-bearing spinels that form adherent, high-density protective surface layers that grow very slowly at operating temperatures.
鉴于压水堆（PWRs）和沸水堆（BWRs）在全球核电站中占有重要比例，这两种类型反应堆的材料问题最受关注。在反应堆的众多材料中，那些承受最极端条件（应力、腐蚀和辐射）的材料对于保障电站安全和可靠性最为关键。图 1 展示了压水堆一回路和二回路的主要部件示意图[10]。一回路中的高压水（约 15.5 MPa）在约 275 °C 下进入反应堆堆芯，从堆芯吸收热量，出口温度约为 325 °C，然后通过蒸汽发生器中的 U 形管将热量传递给低压水。这些水变成蒸汽驱动汽轮机，随后冷凝并循环使用。图 1 还列出了贯穿一回路和二回路所使用的合金，这些合金都与高温水接触，并承受显著的机械应力。反应堆容器内部（包括）的合金也受到不同水平的辐射，这会导致冷却水的位移损伤和辐射分解。 主要的压力边界组件（反应堆压力容器、稳压器、蒸汽发生器、蒸汽管道、汽轮机和冷凝器）由低碳钢或低合金钢制成。奥氏体不锈钢（304、304L、316、316L、321、347）主导着核心结构材料，同时也用于反应堆压力容器和稳压器内表面的包覆（308SS 和 309SS）。弹簧和紧固件等高强度组件由镍基合金制成。容器贯穿件和蒸汽发生器管由镍基合金 690（以前是合金 600，发现其抗应力腐蚀开裂性能不足）制成。冷凝器管通常由钛或不锈钢制成。对于核心内部件和蒸汽发生器管选用镍基合金和奥氏体不锈钢，是由于需要在高温下具有良好的耐水腐蚀性。这些合金由于形成了含铬尖晶石，从而形成致密、高密度的保护性表面层，这些表面层在操作温度下生长非常缓慢，因此腐蚀速率较低。
鉴于压水堆（PWRs）和沸水堆（BWRs）在全球核电站中占有重要比例，这两种类型反应堆的材料问题最受关注。在反应堆的众多材料中，那些承受最极端条件（应力、腐蚀和辐射）的材料对于保障电站安全和可靠性最为关键。图 1 展示了压水堆一回路和二回路的主要部件示意图[10]。一回路中的高压水（约 15.5 MPa）在约 275 °C 下进入反应堆堆芯，从堆芯吸收热量，出口温度约为 325 °C，然后通过蒸汽发生器中的 U 形管将热量传递给低压水。这些水变成蒸汽驱动汽轮机，随后冷凝并循环使用。图 1 还列出了贯穿一回路和二回路所使用的合金，这些合金都与高温水接触，并承受显著的机械应力。反应堆容器内部（包括）的合金也受到不同水平的辐射，这会导致冷却水的位移损伤和辐射分解。 主要的压力边界组件（反应堆压力容器、稳压器、蒸汽发生器、蒸汽管道、汽轮机和冷凝器）由低碳钢或低合金钢制成。奥氏体不锈钢（304、304L、316、316L、321、347）主导着核心结构材料，同时也用于反应堆压力容器和稳压器内表面的包覆（308SS 和 309SS）。弹簧和紧固件等高强度组件由镍基合金制成。容器贯穿件和蒸汽发生器管由镍基合金 690（以前是合金 600，发现其抗应力腐蚀开裂性能不足）制成。冷凝器管通常由钛或不锈钢制成。对于核心内部件和蒸汽发生器管选用镍基合金和奥氏体不锈钢，是由于需要在高温下具有良好的耐水腐蚀性。这些合金由于形成了含铬尖晶石，从而形成致密、高密度的保护性表面层，这些表面层在操作温度下生长非常缓慢，因此腐蚀速率较低。
Fig. 1. Schematic of the primary and secondary circuits of a pressurized water reactor and materials of construction (courtesy of R.W. Staehle) [10].
图 1. 压水堆的主回路和二次回路示意图及建造材料（版权所有 R.W. Staehle）[10]。
The main difference between PWRs and BWRs is that the latter consists of a single water circuit designed for boiling to occur in the core with steam flowing directly to the turbine, which eliminates the steam generator and pressurizer found in the PWR. The operating temperatures are comparable for both reactor types (∼300 °C), with comparable stress and radiation environments. As such, most of the structural alloys are very similar between the two reactor types. The main difference is in the zirconium alloys used as fuel rod cladding, with BWR fuel cladding optimized for corrosion resistance in higher oxygen potentials and PWR fuel cladding optimized for resistance to hydrogen absorption in the low potential environment of the core. Typical zirconium alloy cladding materials used in BWR and PWR reactors are summarized in Table 2. Differences in oxygen potential result in significant impacts on the stress corrosion degradation of materials throughout the water circuit in both reactor types, as will be discussed in Section 2.1.
压水堆（PWR）和沸水堆（BWR）的主要区别在于后者只有一个水循环回路，该回路设计用于在堆芯中沸腾，蒸汽直接流向汽轮机，从而消除了压水堆中存在的蒸汽发生器和稳压器。两种反应堆类型的运行温度相当（约 300°C），应力与辐射环境也相似。因此，两种反应堆类型的结构合金大部分非常相似。主要区别在于用作燃料棒包壳的锆合金，BWR 燃料包壳针对高氧势环境下的耐腐蚀性进行优化，而 PWR 燃料包壳针对堆芯低势环境下的氢吸收抗性进行优化。BWR 和 PWR 反应堆中常用的锆合金包壳材料总结于表 2。氧势的差异导致两种反应堆类型的水循环回路中材料的应力腐蚀降解产生显著影响，这将在 2.1 节中讨论。
压水堆（PWR）和沸水堆（BWR）的主要区别在于后者只有一个水循环回路，该回路设计用于在堆芯中沸腾，蒸汽直接流向汽轮机，从而消除了压水堆中存在的蒸汽发生器和稳压器。两种反应堆类型的运行温度相当（约 300°C），应力与辐射环境也相似。因此，两种反应堆类型的结构合金大部分非常相似。主要区别在于用作燃料棒包壳的锆合金，BWR 燃料包壳针对高氧势环境下的耐腐蚀性进行优化，而 PWR 燃料包壳针对堆芯低势环境下的氢吸收抗性进行优化。BWR 和 PWR 反应堆中常用的锆合金包壳材料总结于表 2。氧势的差异导致两种反应堆类型的水循环回路中材料的应力腐蚀降解产生显著影响，这将在 2.1 节中讨论。
Table 2. Summary of typical commercial zirconium alloys used as cladding in PWRs and BWRs.
表 2. 用于 PWR 和 BWR 堆芯包壳的典型商用锆合金总结。
| Reactor type 反应堆类型 | Zr alloy composition 锆合金成分 | Thermomechanical treatment 热机械处理 |
|---|---|---|
| BWR | Zircaloy-2 (1.5% Sn–0.15% Fe–0.1% Cr–0.05% Ni) | Recrystallized 再结晶 |
| PWR | Zircaloy-4 (1.5% Sn–0.2% Fe–0.1% Cr) | Cold-worked and stress relief anneal 冷加工和应力消除退火 |
| PWR | ZIRLO (1–2% Nb–1% Sn–0.1% Fe) ZIRLO（1-2% Nb-1% Sn-0.1% Fe） | Quench and temper/stress relief anneal 水淬回火/应力消除退火 |
| PWR | M5 (1% Nb) M5（1% Nb） | Recrystallized 再结晶 |
The last reactor design that is in significant use worldwide is the pressurized heavy water reactor (PHWR), the most prevalent version being the CANDU (CANadian Deuterium Uranium) reactor. This reactor uses heavy water as the moderator and primary coolant, transferring heat to light water via a steam generator. The key characteristic of this reactor is the use of deuterium as a moderator, for which neutron absorption is low enough to permit the use of natural (unenriched) uranium, thus bypassing the need for expensive enrichment facilities. A major difference in materials in this system vs. LWRs is the use of Zr–Nb pressure tubes that house the Zircaloy-clad fuel and the high pressure D2O. These tubes fit into Zircaloy-4 calandria tubes that pass through a thin walled stainless steel calandria vessel, which also contains the low temperature D2O moderator. Thus, zirconium alloys play a larger role as pressure boundary materials in PHWRs than they do in LWRs.
目前全球广泛使用的最后一种反应堆设计是压水重水反应堆（PHWR），其中最常见的版本是加拿大重水铀反应堆（CANDU）。这种反应堆使用重水作为慢化剂和主要冷却剂，通过蒸汽发生器将热量传递给轻水。这种反应堆的关键特性是使用氘作为慢化剂，其中子吸收率足够低，允许使用天然（未浓缩）铀，从而避免了建设昂贵浓缩设施的需求。该系统与轻水反应堆（LWR）在材料方面的主要区别在于使用了锆-铌压力管，这些压力管容纳了包有锆合金的燃料和高压重水。这些压力管安装在穿过薄壁不锈钢压力壳的锆合金-4 热交换管内，该压力壳还包含低温重水慢化剂。因此，锆合金在 PHWR 中作为压力边界材料的作用比在 LWR 中更为重要。
目前全球广泛使用的最后一种反应堆设计是压水重水反应堆（PHWR），其中最常见的版本是加拿大重水铀反应堆（CANDU）。这种反应堆使用重水作为慢化剂和主要冷却剂，通过蒸汽发生器将热量传递给轻水。这种反应堆的关键特性是使用氘作为慢化剂，其中子吸收率足够低，允许使用天然（未浓缩）铀，从而避免了建设昂贵浓缩设施的需求。该系统与轻水反应堆（LWR）在材料方面的主要区别在于使用了锆-铌压力管，这些压力管容纳了包有锆合金的燃料和高压重水。这些压力管安装在穿过薄壁不锈钢压力壳的锆合金-4 热交换管内，该压力壳还包含低温重水慢化剂。因此，锆合金在 PHWR 中作为压力边界材料的作用比在 LWR 中更为重要。
Most reactors in the USA and elsewhere in the world were completed in the 1970s and 1980s, and today the average age of the fleet is over 30 years. Fig. 2 shows the worldwide distribution of nuclear power plants classified by years of commercial operation [11]. Since the original license period in the USA is 40 years, many reactor operators are seeking license renewal to allow them to operate the plants for an additional 20 years. To date, 73 of the 104 operating commercial reactors in the USA have received license extensions with another 13 applications under review, and a key question is how long can these plants be safely, reliably and economically operated. The limiting factor is whether critical materials can continue to maintain their integrity beyond 60 years [5]. These materials include reactor components, concrete, cables and buried piping. So the lifetime of the current reactor fleet is ultimately governed by the performance of materials.
美国及其他国家的多数反应堆建于 20 世纪 70 年代和 80 年代，如今该机组的平均年龄已超过 30 年。图 2 展示了按商业运营年份分类的全球核电站分布情况[11]。由于美国的原始许可证期限为 40 年，许多反应堆运营商正寻求许可证续期，以允许他们再将机组运行 20 年。迄今为止，美国 104 台正在运行的商业反应堆中有 73 台已获得许可证延期，另有 13 个申请正在审查中，一个关键问题是这些机组能够安全、可靠和经济地运行多长时间。限制因素在于关键材料能否在 60 年后继续保持其完整性[5]。这些材料包括反应堆部件、混凝土、电缆和埋地管道。因此，当前反应堆机组的寿命最终取决于材料性能。
美国及其他国家的多数反应堆建于 20 世纪 70 年代和 80 年代，如今该机组的平均年龄已超过 30 年。图 2 展示了按商业运营年份分类的全球核电站分布情况[11]。由于美国的原始许可证期限为 40 年，许多反应堆运营商正寻求许可证续期，以允许他们再将机组运行 20 年。迄今为止，美国 104 台正在运行的商业反应堆中有 73 台已获得许可证延期，另有 13 个申请正在审查中，一个关键问题是这些机组能够安全、可靠和经济地运行多长时间。限制因素在于关键材料能否在 60 年后继续保持其完整性[5]。这些材料包括反应堆部件、混凝土、电缆和埋地管道。因此，当前反应堆机组的寿命最终取决于材料性能。
Fig. 2. Age distribution of the world’s commercial nuclear power reactors as of December 2011 [11].
图 2. 截至 2011 年 12 月的世界商业核反应堆年龄分布[11]。
1.2. Major materials degradation modes in nuclear energy systems
1.2. 核能系统中的主要材料退化模式
In addition to satisfying standard materials design criteria based on tensile properties, thermal creep, cyclic fatigue and creep-fatigue, structural materials for current and proposed future nuclear energy systems must provide adequate resistance to two additional overarching environmental degradation phenomena: radiation damage and chemical compatibility. Since the chemical compatibility issues (corrosion, stress corrosion cracking, etc.) are largely dependent on the specific coolant and engineering application, these issues are discussed in the relevant sections on light-water reactors (2.2.2) and advanced reactor concepts (3.2 and 4).
除了要满足基于拉伸性能、热蠕变、循环疲劳和蠕变疲劳的标准材料设计标准外，当前和拟议的未来核能系统的结构材料还必须提供足够的抵抗两种额外的总体环境退化现象的能力：辐射损伤和化学相容性。由于化学相容性问题（腐蚀、应力腐蚀开裂等）在很大程度上取决于特定的冷却剂和工程应用，这些问题将在轻水反应堆（2.2.2）和先进反应堆概念（3.2 和 4）的相关章节中进行讨论。
除了要满足基于拉伸性能、热蠕变、循环疲劳和蠕变疲劳的标准材料设计标准外，当前和拟议的未来核能系统的结构材料还必须提供足够的抵抗两种额外的总体环境退化现象的能力：辐射损伤和化学相容性。由于化学相容性问题（腐蚀、应力腐蚀开裂等）在很大程度上取决于特定的冷却剂和工程应用，这些问题将在轻水反应堆（2.2.2）和先进反应堆概念（3.2 和 4）的相关章节中进行讨论。
There are five key bulk radiation degradation effects (low temperature radiation hardening and embrittlement; radiation-induced and -modified solute segregation and phase stability (including amorphization); irradiation creep; void swelling; and high-temperature helium embrittlement) [8], [12], [13], [14], [15], [16], and a multitude of corrosion and stress corrosion cracking effects in water-cooled reactors [13], [17], [18], [19], [20], [21], [22] and proposed advanced reactors utilizing other coolants [23], [24], [25], [26] (including irradiation-assisted phenomena) that can have a huge impact on the performance of structural materials in nuclear energy systems. The amount of radiation damage produced in materials from exposure to neutrons created by the nuclear energy reactions is quantified by the international standardized parameter [27], [28] of displacements per atom (dpa); a displacement damage value of 1 dpa means that, on average, each atom has been displaced from its lattice site once.
存在五种主要的体辐射退化效应（低温辐射硬化和脆化；辐射诱导和改性溶质偏析及相稳定性（包括非晶化）；辐照蠕变；空位肿胀；以及高温氦脆化）[8], [12], [13], [14], [15], [16]，以及水冷反应堆中大量的腐蚀和应力腐蚀开裂效应[13], [17], [18], [19], [20], [21], [22]，以及利用其他冷却剂的先进反应堆[23], [24], [25], [26]（包括辐照辅助现象），这些都会对核能系统结构材料的性能产生巨大影响。材料因核能反应产生的中子辐照而造成的辐射损伤量，由国际标准化参数位移原子数（dpa）进行量化[27], [28]；1 dpa 的位移损伤值意味着，平均每个原子从其晶格位置移动过一次。
存在五种主要的体辐射退化效应（低温辐射硬化和脆化；辐射诱导和改性溶质偏析及相稳定性（包括非晶化）；辐照蠕变；空位肿胀；以及高温氦脆化）[8], [12], [13], [14], [15], [16]，以及水冷反应堆中大量的腐蚀和应力腐蚀开裂效应[13], [17], [18], [19], [20], [21], [22]，以及利用其他冷却剂的先进反应堆[23], [24], [25], [26]（包括辐照辅助现象），这些都会对核能系统结构材料的性能产生巨大影响。材料因核能反应产生的中子辐照而造成的辐射损伤量，由国际标准化参数位移原子数（dpa）进行量化[27], [28]；1 dpa 的位移损伤值意味着，平均每个原子从其晶格位置移动过一次。
Neutron irradiation can produce pronounced hardening at low and intermediate irradiation temperatures due the production of high densities of nanoscale defect clusters (dislocation loops, helium bubbles, etc.), which serve as obstacles to dislocation motion. This hardening is generally accompanied by a reduction in tensile elongation and fracture toughness. The radiation hardening and reductions in elongation and fracture toughness typically emerge at damage levels above ∼0.1 dpa and are generally most pronounced for homologous irradiation temperatures below 0.35TM, where TM is the absolute melting temperature [26], [29], [30], [31], [32], [33], [34], [35]. Fig. 3 shows an example of the effect of moderate neutron displacement damage levels on the engineering stress–strain curve for austenitic stainless steel [36] and a 8–9% Cr-tempered martensitic steel [35] at 250 °C. Both materials exhibit significant radiation-induced increases in yield and ultimate tensile stress, large reductions in elongation (particularly uniform elongation) and decreased strain hardening capacity. The reductions in elongation and strain hardening capacity have been attributed to flow localization (e.g. dislocation channeling) [37], [38], [39], [40], [41], [42], [43], [44] and strain hardening exhaustion [29], [30], [31] mechanisms. In addition to the decreased elongation, neutron irradiation at low temperature also generally produces a decrease in fracture toughness. Fig. 4 summarizes some of the fracture toughness data for Types 304 and 316 austenitic stainless steels following irradiation at LWR-relevant conditions near 250–350 °C [32], [36], [45], [46], [47], [48]. The fracture toughness decreases rapidly with increasing irradiation dose, and approaches a value near 50 MPa m1/2 after 5–10 dpa. The reduction in fracture toughness can be of particular concern for body-centered cubic materials such as ferritic/martensitic steels if the ductile to brittle transition temperature is shifted to temperatures above cold or warm standby temperatures. The potential for neutron radiation-induced embrittlement of reactor pressure vessel steels has been intensively investigated due to its importance for public safety [49].
中子辐照会在低温和中温辐照时产生明显的硬化现象，这是由于形成了高密度的纳米级缺陷团簇（位错环、氦气泡等），这些团簇阻碍了位错运动。这种硬化通常伴随着抗拉延展性和断裂韧性的降低。辐照硬化和延展性及断裂韧性的降低通常出现在损伤水平高于约 0.1 dpa 时，并且对于同源辐照温度低于 0.35T M 时最为显著，其中 T M 是绝对熔点[26], [29], [30], [31], [32], [33], [34], [35]。图 3 展示了中子位移损伤的适度水平对 250°C 时奥氏体不锈钢[36]和 8-9%铬回火马氏体钢[35]的工程应力-应变曲线的影响。两种材料均表现出辐照引起的屈服应力和抗拉强度的显著增加，延展性大幅降低（尤其是均匀延展性）以及应变硬化能力的下降。 延伸率和应变硬化能力的降低归因于流动局部化（例如位错通道化）[37], [38], [39], [40], [41], [42], [43], [44] 和应变硬化耗尽 [29], [30], [31] 机制。除了延伸率降低外，低温中子辐照通常还会导致断裂韧性下降。图 4 总结了在 250–350 °C 附近、类似轻水堆辐照条件下，304 和 316 奥氏体不锈钢的某些断裂韧性数据 [32], [36], [45], [46], [47], [48]。断裂韧性随辐照剂量的增加而迅速下降，在 5–10 dpa 后接近 50 MPa m 1/2 的值。如果延脆转变温度升高至冷备或热备温度以上，断裂韧性的降低对于体心立方材料（如铁素体/马氏体钢）可能特别值得关注。由于其对公共安全的重要性 [49]，反应堆压力容器钢的中子辐射诱变脆化问题已得到广泛研究。
中子辐照会在低温和中温辐照时产生明显的硬化现象，这是由于形成了高密度的纳米级缺陷团簇（位错环、氦气泡等），这些团簇阻碍了位错运动。这种硬化通常伴随着抗拉延展性和断裂韧性的降低。辐照硬化和延展性及断裂韧性的降低通常出现在损伤水平高于约 0.1 dpa 时，并且对于同源辐照温度低于 0.35T M 时最为显著，其中 T M 是绝对熔点[26], [29], [30], [31], [32], [33], [34], [35]。图 3 展示了中子位移损伤的适度水平对 250°C 时奥氏体不锈钢[36]和 8-9%铬回火马氏体钢[35]的工程应力-应变曲线的影响。两种材料均表现出辐照引起的屈服应力和抗拉强度的显著增加，延展性大幅降低（尤其是均匀延展性）以及应变硬化能力的下降。 延伸率和应变硬化能力的降低归因于流动局部化（例如位错通道化）[37], [38], [39], [40], [41], [42], [43], [44] 和应变硬化耗尽 [29], [30], [31] 机制。除了延伸率降低外，低温中子辐照通常还会导致断裂韧性下降。图 4 总结了在 250–350 °C 附近、类似轻水堆辐照条件下，304 和 316 奥氏体不锈钢的某些断裂韧性数据 [32], [36], [45], [46], [47], [48]。断裂韧性随辐照剂量的增加而迅速下降，在 5–10 dpa 后接近 50 MPa m 1/2 的值。如果延脆转变温度升高至冷备或热备温度以上，断裂韧性的降低对于体心立方材料（如铁素体/马氏体钢）可能特别值得关注。由于其对公共安全的重要性 [49]，反应堆压力容器钢的中子辐射诱变脆化问题已得到广泛研究。
Fig. 3. Effect of neutron irradiation to 3 dpa on the engineering stress–strain curves for (a) solution annealed Type 316LN austenitic steel and (b) 8–9% Cr-tempered martensitic steel at 250 °C (based on Refs. [36] and [35], respectively).
图 3. 中子辐照至 3 dpa 对 250 °C 时 (a) 溶解退火 316LN 奥氏体钢和 (b) 8–9% Cr 回火马氏体钢的工程应力–应变曲线的影响（分别基于参考文献 [36] 和 [35]）。
Fig. 4. Fracture toughness of Types 304 and 316 austenitic stainless steels following irradiation at LWR-relevant conditions near 250–350 °C [32], [36], [45], [46], [47], [48].
图 4. 在 250–350 °C 附近 LWR 相关条件下辐照后的 304 和 316 奥氏体不锈钢的断裂韧性 [32], [36], [45], [46], [47], [48]。
At intermediate temperatures (homologous temperatures >0.3TM), the increased mobility of the radiation defects produces a diverse range of potential microstructural evolutions. The most important radiation degradation phenomena at intermediate temperatures are radiation-induced solute segregation (and associated radiation-induced or -modified precipitation), void swelling, irradiation creep and anisotropic growth. Fig. 5 summarizes the numerous radiation-induced phases that can be induced in initially single-phase austenitic stainless steel as a result of localized radiation-induced solute segregation processes during neutron irradiation [50]. Initial investigations indicated that radiation-induced precipitation was limited to temperatures above 400 °C [51], [52], but recent long-term experiments have observed radiation-induced precipitation in austenitic stainless steel for temperatures as low as 300 °C [50]. Void swelling (due to nucleation and growth of the supersaturation of vacancies produced by irradiation) is characterized by an initial low-swelling transient regime at low doses (during the void nucleation and initial growth phase), followed by a steady-state swelling regime where the volumetric swelling increase is proportional to the dose [12], [16], [53], [54], [55]. Typical post-transient steady-state swelling rates in irradiated metals are ∼0.2–1% dpa−1, which would produce unacceptable volumetric swelling in structural components exposed to high neutron doses. Therefore, research has focused on identifying mechanisms that extend the low-swelling transient regime and delay the onset of the steady-state swelling regime [55], [56]. Irradiation creep [12], [53], [57], [58], [59], [60] and irradiation growth [58], [59], [60], [61] can cause substantial dimensional changes in addition to changes due to void swelling. Irradiation growth is mainly an issue in anisotropic crystallographic systems, such as hexagonal close-packed materials; for this phenomenon, volume is conserved but pronounced anisotropic expansion in one crystallographic direction (and shrinkage in another direction) can occur due to preferential nucleation of defect clusters, such as dislocation loops, on certain crystallographic habit planes. Materials for nuclear energy systems that exhibit irradiation growth include graphite and pure metals or alloys based on zirconium and beryllium. The amount of deformation from irradiation creep is typically proportional to the applied stress and irradiation exposure, with a steady-state creep compliance coefficient of 0.5 to 1 × 10−6 MPa−1 dpa−1 for ferritic and austenitic steels, respectively [12]. Another consequence of irradiation creep is that it can induce undesirable stress relaxation of bolts or springs. Fig. 6 shows the measured stress relaxation for neutron irradiated Inconel X750 springs [62]. Nearly complete relaxation of the initially applied stress on the springs occurred after an irradiation dose of ∼20 dpa at ∼400 °C.
在中间温度（同源温度>0.3T M ）下，辐射缺陷的迁移率增加会导致多种潜在的微观结构演变。中间温度下最重要的辐射退化现象包括辐射诱导的溶质偏析（及相关联的辐射诱导或改性沉淀）、空洞肿胀、辐照蠕变和各向异性生长。图 5 总结了在快中子辐照过程中，由于局部辐射诱导的溶质偏析过程，在初始单相奥氏体不锈钢中可能诱导的众多辐射诱导相[50]。早期研究表明，辐射诱导的沉淀仅限于 400 °C 以上的温度[51],[52]，但最近的长期实验观察到在 300 °C 的奥氏体不锈钢中出现了辐射诱导的沉淀[50]。 空位肿胀（由辐照产生的过饱和空位的成核和生长引起）的特点是：在低剂量下（在空位成核和初始生长阶段）呈现初始低肿胀的瞬态阶段，随后进入稳态肿胀阶段，其中体积肿胀的增加与剂量成正比[12], [16], [53], [54], [55]。辐照金属中典型的瞬态后稳态肿胀率约为 0.2–1% dpa −1 ，这对暴露于高中子剂量的结构部件来说会产生不可接受的体积肿胀。因此，研究重点在于识别能够延长低肿胀瞬态阶段并延缓稳态肿胀阶段开始的机制[55], [56]。辐照蠕变[12], [53], [57], [58], [59], [60]和辐照生长[58], [59], [60], [61]除了引起空位肿胀的变化外，还会导致显著的尺寸变化。 辐照生长主要存在于各向异性晶体学体系中，例如密排六方材料；对于这种现象，体积保持不变，但由于缺陷团簇（如位错环）在特定晶体学晶面发生优先形核，导致在某一晶体学方向上出现显著的各向异性膨胀（而在另一方向上发生收缩）。核能系统中的材料，如石墨以及基于锆和铍的纯金属或合金，会表现出辐照生长。辐照蠕变引起的变形量通常与施加的应力和辐照剂量成正比，对于铁素体和奥氏体钢，稳态蠕变顺应系数分别为 0.5 至 1×10^2 MPa^-1 dpa^-1 [12]。辐照蠕变的另一个后果是它可能引起螺栓或弹簧的不希望应力弛豫。图 6 显示了中子辐照后的 Inconel X750 弹簧的测量应力弛豫情况[62]。在约 400°C 下，经过约 20 dpa 的辐照剂量后，弹簧上初始施加的应力几乎完全弛豫。
在中间温度（同源温度>0.3T M ）下，辐射缺陷的迁移率增加会导致多种潜在的微观结构演变。中间温度下最重要的辐射退化现象包括辐射诱导的溶质偏析（及相关联的辐射诱导或改性沉淀）、空洞肿胀、辐照蠕变和各向异性生长。图 5 总结了在快中子辐照过程中，由于局部辐射诱导的溶质偏析过程，在初始单相奥氏体不锈钢中可能诱导的众多辐射诱导相[50]。早期研究表明，辐射诱导的沉淀仅限于 400 °C 以上的温度[51],[52]，但最近的长期实验观察到在 300 °C 的奥氏体不锈钢中出现了辐射诱导的沉淀[50]。 空位肿胀（由辐照产生的过饱和空位的成核和生长引起）的特点是：在低剂量下（在空位成核和初始生长阶段）呈现初始低肿胀的瞬态阶段，随后进入稳态肿胀阶段，其中体积肿胀的增加与剂量成正比[12], [16], [53], [54], [55]。辐照金属中典型的瞬态后稳态肿胀率约为 0.2–1% dpa −1 ，这对暴露于高中子剂量的结构部件来说会产生不可接受的体积肿胀。因此，研究重点在于识别能够延长低肿胀瞬态阶段并延缓稳态肿胀阶段开始的机制[55], [56]。辐照蠕变[12], [53], [57], [58], [59], [60]和辐照生长[58], [59], [60], [61]除了引起空位肿胀的变化外，还会导致显著的尺寸变化。 辐照生长主要存在于各向异性晶体学体系中，例如密排六方材料；对于这种现象，体积保持不变，但由于缺陷团簇（如位错环）在特定晶体学晶面发生优先形核，导致在某一晶体学方向上出现显著的各向异性膨胀（而在另一方向上发生收缩）。核能系统中的材料，如石墨以及基于锆和铍的纯金属或合金，会表现出辐照生长。辐照蠕变引起的变形量通常与施加的应力和辐照剂量成正比，对于铁素体和奥氏体钢，稳态蠕变顺应系数分别为 0.5 至 1×10^2 MPa^-1 dpa^-1 [12]。辐照蠕变的另一个后果是它可能引起螺栓或弹簧的不希望应力弛豫。图 6 显示了中子辐照后的 Inconel X750 弹簧的测量应力弛豫情况[62]。在约 400°C 下，经过约 20 dpa 的辐照剂量后，弹簧上初始施加的应力几乎完全弛豫。
Fig. 5. Precipitate phases observed in Type 316 austenitic stainless steel after neutron irradiation as a function of temperature and dose. Partially shaded data points at temperatures <400 °C denote the presence of γ′ phase and solid data points are for either G and related phases or an unidentified phase [50].
图 5. 在不同温度和剂量下观察到的 316 型奥氏体不锈钢中子辐照后的沉淀相。温度低于 400 °C 的部分阴影数据点表示存在 γ′ 相，实心数据点表示 G 相及相关相或未识别相 [50]。
Fig. 6. Stress relaxation (normalized to the initial applied stress) for Inconel X750 springs irradiated in the EBR-II fast fission reactor [62].
图 6. 在 EBR-II 快中子增殖堆中辐照的 Inconel X750 弹簧的应力弛豫（相对于初始施加应力进行归一化）[62]。
At high temperatures (above 0.5–0.6TM) the efficient annealing of lattice defects produces recovery of most of the radiation damage. One notable exception is associated with the transmutant He produced from (n, α) reactions within the material. The helium can diffuse to grain boundaries, where it can form large bubbles that weaken the grain boundary strength and cause dramatic reductions in the total elongation [63], [64], [65], [66]. This phenomenon of high-temperature helium embrittlement may restrict the upper operating temperature of materials in nuclear energy systems to temperatures significantly lower than what would be established by thermal creep strength considerations.
在高温（高于 0.5–0.6T M ）下，晶格缺陷的有效退火会导致大部分辐射损伤的恢复。一个值得注意的例外是与材料内部 (n, α) 反应产生的氦同位素有关。氦可以扩散到晶界，在那里它可以形成大气泡，削弱晶界强度并导致总延伸率显著降低 [63], [64], [65], [66]。这种高温氦脆化现象可能会将核能系统中材料的最高工作温度限制在热蠕变强度考虑所确定的温度显著以下。
在高温（高于 0.5–0.6T M ）下，晶格缺陷的有效退火会导致大部分辐射损伤的恢复。一个值得注意的例外是与材料内部 (n, α) 反应产生的氦同位素有关。氦可以扩散到晶界，在那里它可以形成大气泡，削弱晶界强度并导致总延伸率显著降低 [63], [64], [65], [66]。这种高温氦脆化现象可能会将核能系统中材料的最高工作温度限制在热蠕变强度考虑所确定的温度显著以下。
2. Materials challenges in current commercial fission reactors
2. 当前商业裂变反应堆中的材料挑战
2.1. Operating environment for materials in existing LWRs
2.1. 现有压水堆中材料的运行环境
Materials in LWRs are exposed to a variety of conditions. In the following, the operating environment for normal, extended life and transient conditions are summarized. Used fuel disposition issues, while important, are not discussed in this paper.
压水堆中的材料暴露于多种条件。以下总结了正常、延寿和瞬态条件下的运行环境。乏燃料处置问题虽然重要，但本文不作讨论。
压水堆中的材料暴露于多种条件。以下总结了正常、延寿和瞬态条件下的运行环境。乏燃料处置问题虽然重要，但本文不作讨论。
2.1.1. LWR materials under normal operating conditions
2.1.1. 正常运行条件下压水堆材料
Core materials include both fuel materials and structural components. The fuel consists of UO2 pellets in the shape of right circular cylinders with length and diameter of approximately 1 cm each, loaded into 3–4 m long zirconium alloy fuel tubes (cladding), which are grouped into fuel assemblies containing control rods or blades. In BWRs, the assemblies generally contain approximately 100 fuel rods arranged in a square array. Each BWR assembly is encased in a square zirconium alloy tube or fuel channel approximately 12 cm on a side. The control rods are clustered into cruciform-shaped control blades that pass between assemblies clustered in groups of four, as illustrated in Fig. 7. There are typically 700–800 fuel assemblies in a BWR. PWRs contain fewer (∼200), but larger (21 cm on a side) assemblies, containing up to 300 fuel rods of slightly smaller diameter than those in BWRs. The PWR control rods are distributed throughout the square lattice and are connected to each other to form a control rod cluster, as shown in Fig. 8. In both cases, control rods consist of stainless steel tubes filled with boron carbide for neutron absorption. There is no channel box around a PWR assembly, so cross-flow of water between assemblies is possible. The PWR and BWR cores are typically operated nonstop for 18–24 months between refueling operations. Low and intermediate burn-up fuel assemblies are typically moved to different positions in the core during refueling outages to provide optimized fuel management, with a total core residence period of 3–4 fuel cycles (i.e. a third to a quarter of the fuel assemblies are removed each refueling cycle) until they achieve typical cumulative burn-up levels of ∼40–60 GW days per metric ton of uranium (GWd MTU−1), corresponding to fissions in ∼4.2–6.4% of the original uranium atoms.
核心材料包括燃料材料和结构部件。燃料由形状为直圆柱体的 UO₂燃料棒组成，长度和直径均约为 1 厘米，装入 3-4 米长的锆合金燃料管（包壳）中，这些燃料管被组合成包含控制棒或控制叶片的燃料组件。在压水堆（BWR）中，组件通常包含约 100 根燃料棒，排列成方形阵列。每个 BWR 组件被封装在边长约 12 厘米的方形锆合金管或燃料通道中。控制棒被集合成十字形控制叶片，这些叶片穿过由四组组件组成的组件群，如图 7 所示。BWR 中通常有 700-800 个燃料组件。压水堆（PWR）的组件较少（约 200 个），但尺寸更大（边长约 21 厘米），包含多达 300 根直径略小于 BWR 中燃料棒的燃料棒。PWR 的控制棒分布在整个方形晶格中，并相互连接形成控制棒簇，如图 8 所示。在这两种情况下，控制棒均由填充有碳化硼以吸收中子的不锈钢管制成。 压水堆组件周围没有通道箱，因此组件之间的水流是可能的。压水堆和沸水堆堆芯通常在燃料加料操作之间连续运行 18-24 个月。低放和中放燃料组件在燃料加料停堆期间通常会被移动到堆芯的不同位置，以实现优化的燃料管理，整个堆芯的停留时间为 3-4 个燃料循环（即每个燃料加料循环有三分之一到四分之一的燃料组件被移出），直到它们达到典型的累积燃耗水平，约为每吨铀 40-60 吉瓦天（GWd MTU −1 ），相当于原始铀原子中约 4.2-6.4%的裂变。
核心材料包括燃料材料和结构部件。燃料由形状为直圆柱体的 UO₂燃料棒组成，长度和直径均约为 1 厘米，装入 3-4 米长的锆合金燃料管（包壳）中，这些燃料管被组合成包含控制棒或控制叶片的燃料组件。在压水堆（BWR）中，组件通常包含约 100 根燃料棒，排列成方形阵列。每个 BWR 组件被封装在边长约 12 厘米的方形锆合金管或燃料通道中。控制棒被集合成十字形控制叶片，这些叶片穿过由四组组件组成的组件群，如图 7 所示。BWR 中通常有 700-800 个燃料组件。压水堆（PWR）的组件较少（约 200 个），但尺寸更大（边长约 21 厘米），包含多达 300 根直径略小于 BWR 中燃料棒的燃料棒。PWR 的控制棒分布在整个方形晶格中，并相互连接形成控制棒簇，如图 8 所示。在这两种情况下，控制棒均由填充有碳化硼以吸收中子的不锈钢管制成。 压水堆组件周围没有通道箱，因此组件之间的水流是可能的。压水堆和沸水堆堆芯通常在燃料加料操作之间连续运行 18-24 个月。低放和中放燃料组件在燃料加料停堆期间通常会被移动到堆芯的不同位置，以实现优化的燃料管理，整个堆芯的停留时间为 3-4 个燃料循环（即每个燃料加料循环有三分之一到四分之一的燃料组件被移出），直到它们达到典型的累积燃耗水平，约为每吨铀 40-60 吉瓦天（GWd MTU −1 ），相当于原始铀原子中约 4.2-6.4%的裂变。
Fig. 7. Fuel assemblies and control blade used in a boiling water reactor (image credit GE).
图 7. 沸水堆中使用的燃料组件和控制叶片（图片版权 GE）。
Fig. 8. Fuel assembly and control rod cluster used in a pressurized water reactor (image credits: Commissariat à l′énergie atomique and Westinghouse).
图 8. 压水堆中使用的燃料组件和控制棒簇（图片来源：法国原子能委员会和西屋公司）。
Non-fuel core components consist of major structures such as the core shroud (BWR) or the baffle–former assembly (PWR), and smaller components such as bolts, springs, support pins and clips. The core shroud in a BWR is a cylindrical barrel, open at both ends, that surrounds the fuel assemblies. Water from the condenser mixes with water recirculated from the core between the shroud near the top of the vessel and is channeled down the annulus formed by the shroud and the reactor pressure vessel (RPV), then up through the fuel assemblies. In a PWR, the baffle–former assembly plays a similar role in forcing the incoming water down the annulus formed between the baffle–former assembly and the RPV inner diameter and up through the fuel channels to remove heat from fission. Moving out from the core, additional key components are the control rod drive mechanisms and housing, and vessel head penetrations consisting of welded austenitic stainless steel or nickel-base nozzles in the reactor head (PWR) or bottom (BWR), and the RPV.
非燃料核心组件包括主要结构，如沸水堆的核心屏蔽（BWR）或压水堆的挡板-形成组件（PWR），以及较小的组件，如螺栓、弹簧、支撑销和夹子。在沸水堆中，核心屏蔽是一个两端开口的圆柱形筒，包围着燃料组件。来自冷凝器的水与从核心中循环回的水混合，在靠近容器的顶部屏蔽附近混合，然后通过由屏蔽和反应堆压力容器（RPV）形成的环隙向下流动，再通过燃料组件向上流动。在压水堆中，挡板-形成组件在迫使进入的水通过由挡板-形成组件和 RPV 内径形成的环隙向下流动，并通过燃料通道以从裂变中移除热量方面发挥着类似的作用。从核心向外，其他关键组件是控制棒驱动机制和外壳，以及反应堆头部（PWR）或底部（BWR）中由焊接奥氏体不锈钢或镍基喷嘴组成的反应堆头穿透件和 RPV。
非燃料核心组件包括主要结构，如沸水堆的核心屏蔽（BWR）或压水堆的挡板-形成组件（PWR），以及较小的组件，如螺栓、弹簧、支撑销和夹子。在沸水堆中，核心屏蔽是一个两端开口的圆柱形筒，包围着燃料组件。来自冷凝器的水与从核心中循环回的水混合，在靠近容器的顶部屏蔽附近混合，然后通过由屏蔽和反应堆压力容器（RPV）形成的环隙向下流动，再通过燃料组件向上流动。在压水堆中，挡板-形成组件在迫使进入的水通过由挡板-形成组件和 RPV 内径形成的环隙向下流动，并通过燃料通道以从裂变中移除热量方面发挥着类似的作用。从核心向外，其他关键组件是控制棒驱动机制和外壳，以及反应堆头部（PWR）或底部（BWR）中由焊接奥氏体不锈钢或镍基喷嘴组成的反应堆头穿透件和 RPV。
The RPV serves as both a pressure barrier and a containment barrier for the radioactive fission products produced during the nuclear fission reaction, and thus plays a key role in reactor safety. The RPV is typically constructed from carbon and low alloy ferritic steels with 1–2% Mn, 0.5–1% Ni, ∼0.5% Mo and 0.15–0.4% Si [67], and has a typical wall thickness of ∼20 cm. Older LWR pressure vessels were constructed from rolled plates that were welded to form a large cylinder, whereas newer vessels are formed from ring forgings in order to eliminate welds in the vessel “beltline” region closest to the center of the reactor core. The top and bottom heads are usually constructed from low alloy steel forgings, and are welded to the central cylindrical vessel (or bolted and gasketed in the case of the upper head). The internal surface of the RPV is typically clad with 5–10 mm of an austenitic stainless steel to provide corrosion compatibility with the reactor coolant. Multiple penetrations for coolant flow and instrumentation are made through the pressure vessel. The fast neutron flux is three to four orders of magnitude lower at the RPV compared to core internal structures [67], but it is still of sufficient intensity to cause radiation hardening, which could lead to fracture toughness embrittlement. It is important to maintain adequate levels of fracture toughness for a wide variety of operational conditions, including normal operation, cold shutdown for refueling and other maintenance, and postulated transient accident scenarios such as pressurized thermal shock in PWRs that would introduce cold water into the reactor vessel while the vessel is at operating pressure and temperature (creating large thermal stresses and potential for crack propagation).
压力容器（RPV）既作为放射性裂变产物在核裂变反应中产生的压力屏障，也作为包容屏障，因此在反应堆安全中发挥着关键作用。压力容器通常由含 1-2% Mn、0.5-1% Ni、~0.5% Mo 和 0.15-0.4% Si 的碳钢和低合金铁素体钢制成[67]，其典型壁厚约为 20 厘米。较旧的轻水反应堆压力容器由滚制钢板焊接而成，形成一个大圆柱体，而较新的容器则由环锻制成，以消除靠近反应堆核心中心的容器“腰带区”的焊缝。顶部和底部封头通常由低合金钢锻件制成，并焊接到中央圆柱形容器（或上封头采用螺栓和垫片连接）。压力容器的内表面通常覆盖 5-10 毫米的奥氏体不锈钢，以提供与反应堆冷却剂的腐蚀兼容性。冷却剂流动和仪表的多个穿透孔通过压力容器制成。 快中子通量在反应堆压力容器（RPV）处比堆芯内部结构低三个到四个数量级[67]，但它仍然具有足够的强度，足以引起辐照硬化，可能导致断裂韧性脆化。对于包括正常运行、为燃料更换而进行的冷停堆和其他维护，以及假定的瞬态事故场景（如压水堆中的压热冲击，这种冲击会在压力容器处于运行压力和温度时引入冷水，从而产生巨大的热应力并可能引发裂纹扩展）在内的各种操作条件，保持足够的断裂韧性非常重要。
压力容器（RPV）既作为放射性裂变产物在核裂变反应中产生的压力屏障，也作为包容屏障，因此在反应堆安全中发挥着关键作用。压力容器通常由含 1-2% Mn、0.5-1% Ni、~0.5% Mo 和 0.15-0.4% Si 的碳钢和低合金铁素体钢制成[67]，其典型壁厚约为 20 厘米。较旧的轻水反应堆压力容器由滚制钢板焊接而成，形成一个大圆柱体，而较新的容器则由环锻制成，以消除靠近反应堆核心中心的容器“腰带区”的焊缝。顶部和底部封头通常由低合金钢锻件制成，并焊接到中央圆柱形容器（或上封头采用螺栓和垫片连接）。压力容器的内表面通常覆盖 5-10 毫米的奥氏体不锈钢，以提供与反应堆冷却剂的腐蚀兼容性。冷却剂流动和仪表的多个穿透孔通过压力容器制成。 快中子通量在反应堆压力容器（RPV）处比堆芯内部结构低三个到四个数量级[67]，但它仍然具有足够的强度，足以引起辐照硬化，可能导致断裂韧性脆化。对于包括正常运行、为燃料更换而进行的冷停堆和其他维护，以及假定的瞬态事故场景（如压水堆中的压热冲击，这种冲击会在压力容器处于运行压力和温度时引入冷水，从而产生巨大的热应力并可能引发裂纹扩展）在内的各种操作条件，保持足够的断裂韧性非常重要。
Materials utilized in LWR cores must withstand simultaneous application of mechanical stress, neutron irradiation and corrosion due to hot water or steam (see rows 1 and 2 of Table 3). Temperatures of core components are in the range 275–288 °C in BWRs and 290–320 °C in PWRs. The BWR environment is characterized by an electrochemical potential (ECP) in the range of 150 mV relative to the standard hydrogen electrode, or 150 mVSHE, due to a combination of boiling of the water in the core and radiolysis. PWRs operate at a lower potential (<−500 mVSHE) by virtue of the addition of hydrogen at a level of 35 cc of H2 per kg of water (∼3 ppm) to scavenge radiolysis products and lower the corrosion potential. PWR primary water also contains 1000 ppm B as boric acid (H3BO3) added for reactivity control and 2–4 ppm Li as LiOH added for pH control. The lower ECP is better for both corrosion and stress corrosion cracking of core materials and is possible in a PWR due to the lack of boiling. In addition to controlling pH, boron impacts the formation of solid corrosion products (CRUD) and corrosion of fuel cladding, as well as reactivity control of the reactor, and is briefly described in Section 2.2. Stresses on fuel and core components come from a variety of sources, including thermal expansion, high velocity water flow, residual stresses due to welding, and stresses due to radiation-induced volume expansion or distortion.
用于轻水堆芯的材料必须能够承受机械应力、中子辐照以及由于热水或蒸汽引起的腐蚀的共同作用（参见表 3 的第 1 行和第 2 行）。在沸水堆中，堆芯组件的温度范围为 275–288 °C，在压水堆中为 290–320 °C。沸水堆环境的特点是相对于标准氢电极的电化学电位（ECP）在 150 mV 的范围内，或 150 mV SHE ，这是由于堆芯中水的沸腾和辐射分解的共同作用所致。压水堆通过向水中添加氢气（每千克水 35 cc H 2 ，约 3 ppm）来清除辐射分解产物并降低腐蚀电位，从而在较低电位下（<−500 mV SHE ）运行。压水堆的初级水还含有 1000 ppm 的硼，以硼酸（H 3 BO 3 ）的形式添加，用于反应性控制，以及 2–4 ppm 的锂，以氢氧化锂（LiOH）的形式添加，用于 pH 控制。较低的 ECP 对堆芯材料的腐蚀和应力腐蚀开裂都有利，这是由于压水堆中没有沸腾现象而得以实现。 除了控制 pH 值外，硼还会影响固体腐蚀产物（CRUD）的形成、燃料包壳的腐蚀以及反应堆的反应性控制，这些内容在 2.2 节中简要介绍。燃料和核心部件所承受的应力来自多种来源，包括热膨胀、高速水流、焊接残余应力以及辐射引起的体积膨胀或变形应力。
用于轻水堆芯的材料必须能够承受机械应力、中子辐照以及由于热水或蒸汽引起的腐蚀的共同作用（参见表 3 的第 1 行和第 2 行）。在沸水堆中，堆芯组件的温度范围为 275–288 °C，在压水堆中为 290–320 °C。沸水堆环境的特点是相对于标准氢电极的电化学电位（ECP）在 150 mV 的范围内，或 150 mV SHE ，这是由于堆芯中水的沸腾和辐射分解的共同作用所致。压水堆通过向水中添加氢气（每千克水 35 cc H 2 ，约 3 ppm）来清除辐射分解产物并降低腐蚀电位，从而在较低电位下（<−500 mV SHE ）运行。压水堆的初级水还含有 1000 ppm 的硼，以硼酸（H 3 BO 3 ）的形式添加，用于反应性控制，以及 2–4 ppm 的锂，以氢氧化锂（LiOH）的形式添加，用于 pH 控制。较低的 ECP 对堆芯材料的腐蚀和应力腐蚀开裂都有利，这是由于压水堆中没有沸腾现象而得以实现。 除了控制 pH 值外，硼还会影响固体腐蚀产物（CRUD）的形成、燃料包壳的腐蚀以及反应堆的反应性控制，这些内容在 2.2 节中简要介绍。燃料和核心部件所承受的应力来自多种来源，包括热膨胀、高速水流、焊接残余应力以及辐射引起的体积膨胀或变形应力。
Table 3. Reactor core environment and materials for light water reactors and advanced fission reactor concepts [7].
表 3. 轻水反应堆和先进裂变反应堆概念的反应堆核心环境与材料[7].
| System 系统 | Coolant 冷却剂 | Pressure (MPa) 压力（MPa） | Tin/Tout (°C) | Neutron spectrum, maximum dose (dpa) 中子谱，最大剂量（dpa） | Fuel 燃料 | Cladding 包覆层 | Structural materials 结构材料 | |
|---|---|---|---|---|---|---|---|---|
| In-core 堆芯内 | Out-of-core 堆芯外 | |||||||
| Pressurized water reactor – PWR 压水堆 – PWR | Water – single phase 水 – 单相 | 16 | 290/320 | Thermal, ∼80 热，~80 | UO2 (or MOX) UO 2 (或 MOX) | Zirconium alloy 锆合金 | Stainless steels, nickel-based alloys 不锈钢、镍基合金 | Stainless steels, nickel-based alloys 不锈钢、镍基合金 |
| Boiling water reactor – BWR 沸水堆 – BWR | Water – two phase 水——两相 | 7 | 280/288 | Thermal, ∼7 热，~7 | UO2 (or MOX) UO 2 （或 MOX） | Zircaloy 锆合金 | Stainless steels, nickel-based alloys 不锈钢，镍基合金 | Stainless steels, nickel-based alloys 不锈钢，镍基合金 |
| Supercritical water cooled reactor – SCWR 超临界水冷反应堆 – SCWR | Supercritical water 超临界水 | 25 | 290/600 | Thermal, ∼30, fast, ∼70 热，约 30，快，约 70 | UO2 | F-M (12Cr, 9Cr, etc.) (Fe–35Ni–25Cr–0.3Ti), Incoloy 800, ODS, Inconel 690, 625, and 718 F-M（12Cr、9Cr 等）（Fe–35Ni–25Cr–0.3Ti）、Incoloy 800、ODS、Inconel 690、625 和 718 | Same as cladding options, plus low swelling stainless steels 与包壳选项相同，此外还有低膨胀不锈钢 | F-M, low-alloy steels F-M、低合金钢 |
| Very high temperature reactor – VHTR 超高温反应堆 – VHTR | Helium 氦气 | 7 | 600/1000 | Thermal, <20 热能，<20 | UO2, UCO | SiC or ZrC coating and surrounding graphite SiC 或 ZrC 涂层及周围的石墨 | Graphites, PyC, SiC, ZrC, vessel: F-M 石墨、PyC、SiC、ZrC、容器：F-M | Ni-based superalloys, 32Ni–25Cr–20Fe–12.5W–0.05C, Ni–23Cr–18W–0.2C, F-M w/thermal barriers, low-alloy steels 镍基高温合金，32Ni–25Cr–20Fe–12.5W–0.05C，Ni–23Cr–18W–0.2C，F-M 带热障，低合金钢 |
| Gas fast reactor – GFR 气冷快堆 – GFR | Helium, supercritical CO2 氦气，超临界 CO₂ | 7 | 450/850 | Fast, 80 快速，80 | MC | Ceramic 陶瓷 | Refractory metals and alloys, Ceramics, ODS, vessel: F-M 难熔金属和合金，陶瓷，ODS，容器：F-M | Ni-based superalloys, 32Ni–25Cr–20Fe–12.5W–0.05C, Ni–23Cr–18W–0.2C, F-M w/therm barriers 镍基高温合金，32Ni–25Cr–20Fe–12.5W–0.05C，Ni–23Cr–18W–0.2C，F-M 带热障 |
| Sodium fast reactor – SFR 钠冷快堆 – SFR | Sodium 钠 | 0.1 | 370/550 | Fast, 200 快，200 | MOX or U–Pu–Zr or MC or MN MOX 或 U–Pu–Zr 或 MC 或 MN | F-M or F-M ODS F-M 或 F-M ODS | F-M ducts, 316SS grid plate F-M 管道，316SS 网格板 | Ferritics, austenitics 铁素体，奥氏体 |
| Lead fast reactor – LFR 铅快堆 – LFR | Lead or lead–bismuth 铅或铅铋 | 0.1 | 600/800 | Fast, 150 快，150 | MN | High-Si F-M, ODS, ceramics, or refractory alloys 高 Si F-M，ODS，陶瓷或难熔合金 | High-Si austenitics, ceramics, or refractory alloys 高硅奥氏体、陶瓷或耐火合金 | |
| Molten salt reactor – MSR 熔盐反应堆——MSR | Molten salt, for example: FLiNaK 熔盐，例如：FLiNaK | 0.1 | 700/1000 | Thermal, 200 热，200 | Salt 盐 | Not applicable 不适用 | Ceramics, refractory metals, Mo, Ni-alloys, (e.g., INOR-8), graphite, Hastelloy N 陶瓷、难熔金属、钼、镍合金（例如，INOR-8）、石墨、哈斯特洛依镍合金 | High-Mo, Ni-based alloys (e.g., INOR-8) 高钼、镍基合金（例如，INOR-8） |
Abbreviations: F-M, Ferritic–martensitic stainless steels (typically 9–12 wt.% Cr); ODS, oxide dispersion-strengthened steels (typically ferritic–martensitic); MC, mixed carbide (U,Pu)C; MN, mixed nitride (U,Pu)N; MOX, mixed oxide (U,Pu)O2.
缩写：F-M，铁素体-马氏体不锈钢（通常含 9-12 wt.% Cr）；ODS，氧化物弥散强化钢（通常为铁素体-马氏体）；MC，混合碳化物（U,Pu）C；MN，混合氮化物（U,Pu）N；MOX，混合氧化物（U,Pu）O 2 。
缩写：F-M，铁素体-马氏体不锈钢（通常含 9-12 wt.% Cr）；ODS，氧化物弥散强化钢（通常为铁素体-马氏体）；MC，混合碳化物（U,Pu）C；MN，混合氮化物（U,Pu）N；MOX，混合氧化物（U,Pu）O 2 。
The unique environmental element of a reactor core is radiation. Fission results in several different types of radiation that affect materials in different ways. Principal radiation types contributing to material degradation are the fission products, neutrons and gamma rays. Fission products consist of high-energy (∼100 MeV) elements of sizable mass (generally between 90 and 150 atomic mass units) that result directly from the fission process. These elements are created as highly charged ions that deposit their energy within 10 μm of their origin. As such, except for those that are born within this distance of the fuel pellet surface, fission product damage is confined to the fuel. The fission process releases both neutrons and gammas.
反应堆堆芯独特的环境因素是辐射。裂变会产生多种不同类型的辐射，这些辐射以不同方式影响材料。导致材料退化的主要辐射类型是裂变产物、中子和伽马射线。裂变产物由高能（~100 MeV）的较重元素组成（通常质量数在 90 到 150 之间），这些元素直接由裂变过程产生。这些元素以高电荷离子的形式被创建，并在其产生点附近 10 μm 内沉积能量。因此，除了那些在燃料芯块表面附近 10 μm 内产生的裂变产物外，裂变产物损伤仅限于燃料。裂变过程同时释放中子和伽马射线。
反应堆堆芯独特的环境因素是辐射。裂变会产生多种不同类型的辐射，这些辐射以不同方式影响材料。导致材料退化的主要辐射类型是裂变产物、中子和伽马射线。裂变产物由高能（~100 MeV）的较重元素组成（通常质量数在 90 到 150 之间），这些元素直接由裂变过程产生。这些元素以高电荷离子的形式被创建，并在其产生点附近 10 μm 内沉积能量。因此，除了那些在燃料芯块表面附近 10 μm 内产生的裂变产物外，裂变产物损伤仅限于燃料。裂变过程同时释放中子和伽马射线。
Neutrons are created with an energy of ∼2 MeV and slow down via collisions with the coolant and structural components. Neutrons are the primary source of radiation damage to core materials and fuel cladding and assembly components. Typical displacement damage exposures to the fuel cladding (replaced every ∼5 years) are about 15 dpa. The cumulative displacement damage in core internal structures can approach 80 dpa after 40 years. The displacement damage rate decreases rapidly with increasing distance from the core due to neutron moderation (lower neutron energy resulting from energy loss via collisions with the coolant and core materials); the displacement damage levels in the RPV wall for a PWR are typically ∼0.05 dpa after 40 years of operation, and the corresponding damage in a BWR vessel wall can be up to an order of magnitude lower. This displacement damage can result in significant temperature- and dose-dependent changes to the microstructure (formation of dislocation loops, precipitate formation/dissolution, void formation, radiation-induced segregation, etc.) [16], which affects mechanical properties (strength/hardness, ductility, fracture toughness and embrittlement, creep, fatigue). When combined with the environment, high temperature and stress, additional modes of degradation occur such as irradiation-assisted stress corrosion cracking, corrosion fatigue and environmentally enhanced fracture toughness degradation.
中子被产生时具有约 2 MeV 的能量，并通过与冷却剂和结构部件的碰撞而减速。中子是核心材料、燃料包壳和组件辐射损伤的主要来源。燃料包壳（约每 5 年更换一次）的典型位移损伤剂量约为 15 dpa。核心内部结构的累积位移损伤在 40 年后可接近 80 dpa。由于中子慢化（因与冷却剂和核心材料碰撞而能量损失，导致中子能量降低），位移损伤率随距离核心的增加而迅速下降；压水堆的 RPV 壁在运行 40 年后的位移损伤水平通常约为 0.05 dpa，而沸水堆容器壁的相应损伤可低一个数量级。 这种位移损伤会导致微观结构发生显著的温度和剂量依赖性变化（位错环形成、析出物形成/溶解、空洞形成、辐照诱导偏析等）[16]，这会影响力学性能（强度/硬度、延展性、断裂韧性及脆化、蠕变、疲劳）。当与环境和高温、应力结合时，还会发生其他形式的退化，如辐照辅助应力腐蚀开裂、腐蚀疲劳和环境增强的断裂韧性退化。
中子被产生时具有约 2 MeV 的能量，并通过与冷却剂和结构部件的碰撞而减速。中子是核心材料、燃料包壳和组件辐射损伤的主要来源。燃料包壳（约每 5 年更换一次）的典型位移损伤剂量约为 15 dpa。核心内部结构的累积位移损伤在 40 年后可接近 80 dpa。由于中子慢化（因与冷却剂和核心材料碰撞而能量损失，导致中子能量降低），位移损伤率随距离核心的增加而迅速下降；压水堆的 RPV 壁在运行 40 年后的位移损伤水平通常约为 0.05 dpa，而沸水堆容器壁的相应损伤可低一个数量级。 这种位移损伤会导致微观结构发生显著的温度和剂量依赖性变化（位错环形成、析出物形成/溶解、空洞形成、辐照诱导偏析等）[16]，这会影响力学性能（强度/硬度、延展性、断裂韧性及脆化、蠕变、疲劳）。当与环境和高温、应力结合时，还会发生其他形式的退化，如辐照辅助应力腐蚀开裂、腐蚀疲劳和环境增强的断裂韧性退化。
The gamma radiation field is intense and extends throughout the core, aided by (n, γ) reactions in structural components. While atom displacement by gamma rays is of minor consequence, the main importance of gamma rays is in their heating and changes to water chemistry. Gamma heating can elevate temperatures in thicker components close to the fuel (such as baffle–former plates and bolts) by as much as 60 °C above the water temperature. Gamma rays also induce radiolysis of the water and create a number of radicals that elevate the corrosion potential in the core. Corrosion potential is the critical element governing stress corrosion cracking of core materials at elevated temperature.
伽马射线场强度大，贯穿整个反应堆芯，这得益于结构部件中的(n, γ)反应。虽然伽马射线引起的原子位移影响不大，但伽马射线的最主要作用在于其加热效应和水化学性质的改变。伽马加热可以使靠近燃料的较厚部件（如挡板-成型板和螺栓）的温度比水温高出高达 60°C。伽马射线还会引发水的辐射分解，产生多种自由基，从而提高反应堆芯的腐蚀电位。腐蚀电位是控制高温下反应堆材料应力腐蚀开裂的关键因素。
伽马射线场强度大，贯穿整个反应堆芯，这得益于结构部件中的(n, γ)反应。虽然伽马射线引起的原子位移影响不大，但伽马射线的最主要作用在于其加热效应和水化学性质的改变。伽马加热可以使靠近燃料的较厚部件（如挡板-成型板和螺栓）的温度比水温高出高达 60°C。伽马射线还会引发水的辐射分解，产生多种自由基，从而提高反应堆芯的腐蚀电位。腐蚀电位是控制高温下反应堆材料应力腐蚀开裂的关键因素。
Beyond the materials contained within the reactor vessel, the major components affected by the water chemistry environment include piping, turbine rotors and blades, the condenser and, in PWRs, the pressurizer and steam generator. Historically, the main materials degradation problems have been intergranular stress corrosion cracking (IGSCC) of BWR piping and of steam generator tubes and in vessel penetrations in PWRs. IGSCC of BWR pipes and steam lines made of 304 stainless steel was due to a combination of weld-induced residual stresses and sensitization of grain boundaries caused by heat treatment of high carbon steels in the temperature region in which chromium carbides formed rapidly on the boundaries, depleting them of chromium and making them susceptible to attack. IGSCC of Alloy 600 occurred in steam generators, on both the primary and secondary sides, and was driven by a susceptible microstructure and the creation of crevices on the secondary side in which crevice chemistry was favorable for intergranular attack.
除了反应堆压力容器内的材料外，受水化学环境影响的主要部件包括管道、涡轮转子和叶片、冷凝器，以及在压水堆（PWRs）中的稳压器和蒸汽发生器。历史上，主要的材料退化问题包括沸水堆（BWR）管道和蒸汽发生器管的晶间应力腐蚀开裂（IGSCC），以及压水堆（PWRs）压力容器贯穿件的 IGSCC。沸水堆 304 不锈钢管道和蒸汽管道的 IGSCC 是由于焊接残余应力与高温处理导致高碳钢晶界区域迅速形成铬碳化物，使晶界脱铬并使其易受侵蚀的组合作用所致。合金 600 的 IGSCC 发生在蒸汽发生器中，包括一次侧和二次侧，其驱动因素是易感的微观结构和二次侧产生的缝隙，缝隙化学环境有利于晶间侵蚀。
除了反应堆压力容器内的材料外，受水化学环境影响的主要部件包括管道、涡轮转子和叶片、冷凝器，以及在压水堆（PWRs）中的稳压器和蒸汽发生器。历史上，主要的材料退化问题包括沸水堆（BWR）管道和蒸汽发生器管的晶间应力腐蚀开裂（IGSCC），以及压水堆（PWRs）压力容器贯穿件的 IGSCC。沸水堆 304 不锈钢管道和蒸汽管道的 IGSCC 是由于焊接残余应力与高温处理导致高碳钢晶界区域迅速形成铬碳化物，使晶界脱铬并使其易受侵蚀的组合作用所致。合金 600 的 IGSCC 发生在蒸汽发生器中，包括一次侧和二次侧，其驱动因素是易感的微观结构和二次侧产生的缝隙，缝隙化学环境有利于晶间侵蚀。
2.1.2. Life extension and power uprates
2.1.2. 延寿和功率提升
Due to the initial high capital cost for construction of nuclear power plants relative to the cost of the fuel and other operating expenses, the levelized cost of electricity for LWRs is dominated by the amortized original cost of construction; the annualized costs associated with the fuel and operating and maintenance costs in a new nuclear power plant are estimated to contribute about 20% of the levelized cost of nuclear electricity [68]. This factor, along with the high capacity factor of LWRs demonstrated during the past decade, has led to significant interest in extending the operational licenses of nuclear power plants beyond their initial term (typically 40 years). Extension of reactor operating licenses for an additional 20 years means that reactor components will be required to maintain their integrity for a period that is 50% longer than the initial 40-year license. This increase in operational life introduces a wide range of potential materials aging issues that must be considered as part of the renewal license process [5], [69]. While the effect of increased irradiation exposure is dependent on the component, the increase in operating life by 50% means that displacement damage at the bottom of the top guide in a BWR may be >50 dpa, while the shroud will acquire a damage level 100 times lower. High-fluence components such as baffle bolts will reach damage levels exceeding 100 dpa. Fig. 9 shows a rough approximation of the fluence (damage) levels associated with various component failures (top) in both BWRs and PWRs or with microstructure/property changes (bottom) and the impact of a 20-year increase in operating lifetime. Life extension will increase the maximum expected damage level on the components receiving the highest fluence, and it will also elevate the damage level on the balance of components proportionately. Additional life extensions beyond 20 years are also being contemplated and the fluence (damage) level coinciding with three 20-year life extensions is shown for comparison.
由于核电站的初始建设成本相对于燃料和其他运营费用较高，轻水反应堆的平准化电力成本主要受初始建设成本的摊销影响；新核电站的燃料、运营和维护相关年化成本预计约占核电力平准化成本的 20%[68]。这一因素，加上过去十年中轻水反应堆展示的高容量因子，已导致人们显著关注将核电站的运营许可证期限延长至初始期限（通常为 40 年）之外。将反应堆运营许可证再延长 20 年意味着反应堆部件需要保持其完整性，其持续时间比初始 40 年许可证长 50%。这种运营寿命的增加引入了一系列潜在的材料老化问题，这些问题必须在更新许可证过程中予以考虑[5][69]。 虽然增加辐照暴露的影响取决于部件，但运行寿命增加 50%意味着在沸水堆中，顶部导向器底部的位移损伤可能>50 dpa，而屏蔽部件的损伤水平将低 100 倍。像挡板螺栓这样的高通量部件将达到超过 100 dpa 的损伤水平。图 9 大致展示了与沸水堆和压水堆中各种部件失效（顶部）或微观结构/性能变化（底部）相关的通量（损伤）水平，以及运行寿命增加 20 年的影响。寿命延长将增加接收最高通量的部件的最大预期损伤水平，并按比例提高其他部件的损伤水平。此外，还考虑了超过 20 年的额外寿命延长，并显示了与三个 20 年寿命延长相对应的通量（损伤）水平以供比较。
由于核电站的初始建设成本相对于燃料和其他运营费用较高，轻水反应堆的平准化电力成本主要受初始建设成本的摊销影响；新核电站的燃料、运营和维护相关年化成本预计约占核电力平准化成本的 20%[68]。这一因素，加上过去十年中轻水反应堆展示的高容量因子，已导致人们显著关注将核电站的运营许可证期限延长至初始期限（通常为 40 年）之外。将反应堆运营许可证再延长 20 年意味着反应堆部件需要保持其完整性，其持续时间比初始 40 年许可证长 50%。这种运营寿命的增加引入了一系列潜在的材料老化问题，这些问题必须在更新许可证过程中予以考虑[5][69]。 虽然增加辐照暴露的影响取决于部件，但运行寿命增加 50%意味着在沸水堆中，顶部导向器底部的位移损伤可能>50 dpa，而屏蔽部件的损伤水平将低 100 倍。像挡板螺栓这样的高通量部件将达到超过 100 dpa 的损伤水平。图 9 大致展示了与沸水堆和压水堆中各种部件失效（顶部）或微观结构/性能变化（底部）相关的通量（损伤）水平，以及运行寿命增加 20 年的影响。寿命延长将增加接收最高通量的部件的最大预期损伤水平，并按比例提高其他部件的损伤水平。此外，还考虑了超过 20 年的额外寿命延长，并显示了与三个 20 年寿命延长相对应的通量（损伤）水平以供比较。
Fig. 9. Neutron fluence (E > 1 MeV) and corresponding displacement damage levels and the corresponding failure modes or changes in microstructure/properties of BWR and PWR components and materials. Dashed lines refer to the maximum expected fluence/dose in PWR core components after the designated time period.
图 9. 中子注量（E > 1 MeV）以及相应的位移损伤水平和沸水堆（BWR）及压水堆（PWR）组件和材料的相应失效模式或微观结构/性能变化。虚线表示指定时间后压水堆堆芯组件的最大预期注量/剂量。
In addition to life extension activities, a second approach to leveraging the existing capital assets of a nuclear power plant is to make modifications to the operating parameters (e.g. coolant flow rate) and/or changes to existing equipment (e.g. turbines) that enable high power levels to be achieved. These power uprate requests require detailed safety analysis review. Since 1977, a total of 139 power uprates ranging from 0.4 to 18% have been approved by the US Nuclear Regulatory Commission, resulting in 6 GWe additional generating capacity. The major impacts of the power uprate on the reactor materials include slightly higher operating temperature and higher neutron flux and fluence to permanent core internal structures.
除了寿命延长活动外，利用核电站现有固定资产的另一种方法是修改运行参数（例如冷却剂流量）和/或更改现有设备（例如涡轮机），以实现高功率水平。这些提高功率的请求需要进行详细的安全分析审查。自 1977 年以来，美国核管会已批准了 139 次从 0.4%到 18%不等的功率提升，增加了 6 GW 的发电能力。功率提升对反应堆材料的主要影响包括略微提高运行温度以及永久堆芯内部结构的中子通量和注量增加。
除了寿命延长活动外，利用核电站现有固定资产的另一种方法是修改运行参数（例如冷却剂流量）和/或更改现有设备（例如涡轮机），以实现高功率水平。这些提高功率的请求需要进行详细的安全分析审查。自 1977 年以来，美国核管会已批准了 139 次从 0.4%到 18%不等的功率提升，增加了 6 GW 的发电能力。功率提升对反应堆材料的主要影响包括略微提高运行温度以及永久堆芯内部结构的中子通量和注量增加。
A potential consequence of higher damage levels from power uprates and/or life extension is the appearance of new or unanticipated degradation modes. One such mode is void swelling. The formation and growth of voids was discovered in the development of materials for fast reactors [70], where core temperatures range from 350 to 550 °C and damage rates are approximately 10 times that in thermal reactors. Since void formation in stainless steels under fast reactor irradiation conditions is pronounced only for irradiation temperatures of 400–650 °C [52], void swelling was not initially considered an issue for LWR core materials, where maximum core component temperatures are below this temperature range. However, as predicted by void swelling models [71] and recently confirmed by long-term LWR irradiation experiments [72], lower damage rates can be more damaging than high dose rates as they induce void formation at lower fluences and also extend the void swelling regime to lower temperatures. Further, as noted in Section 2.1.1, gamma heating causes the temperature in thicker components to increase up to 380 °C, pulling them into the range where void swelling readily occurs in austenitic stainless steels at LWR-relevant dose rates [52], [73]. Baffle–former bolts in PWRs have developed voids at damage levels as low as 7.5 dpa [13]. If the plates fastened by the bolt also swell, they will apply a stress to the bolts, creating conditions that are ripe for irradiation assisted stress corrosion cracking (IASCC). To date, there have been several instances of cracked baffle–former bolts caused by IASCC [20].
从功率提升和/或寿命延长导致更高的损伤水平可能产生的一个潜在后果是新出现或未预料到的退化模式。其中一种模式是空位肿胀。在快堆材料开发过程中发现了空位的形成和生长[70]，快堆的反应堆芯温度范围为 350 至 550°C，损伤速率约为热堆的 10 倍。由于在快堆辐照条件下，不锈钢中的空位形成仅在辐照温度为 400–650°C 时才显著[52]，因此空位肿胀最初不被认为是压水堆（LWR）反应堆芯材料的问题，因为反应堆芯组件的最高温度低于这一温度范围。然而，正如空位肿胀模型[71]所预测的，并且最近由长期压水堆辐照实验[72]所证实，较低的损伤速率可能比高剂量率更具破坏性，因为它们在较低的注量下诱导空位形成，并将空位肿胀范围扩展到较低的温度。 此外，如第 2.1.1 节所述，伽马加热会导致较厚部件的温度升高至 380°C，将其推入奥氏体不锈钢在核电站相关剂量率下容易发生空位肿胀的范围内[52][73]。在压水堆中，挡板-成型螺栓在损伤水平低至 7.5 dpa 时就会产生空位[13]。如果螺栓固定的板也发生肿胀，它们会对螺栓施加应力，从而为辐照辅助应力腐蚀开裂（IASCC）创造条件。迄今为止，已有几例因 IASCC 导致的挡板-成型螺栓开裂的实例[20]。
从功率提升和/或寿命延长导致更高的损伤水平可能产生的一个潜在后果是新出现或未预料到的退化模式。其中一种模式是空位肿胀。在快堆材料开发过程中发现了空位的形成和生长[70]，快堆的反应堆芯温度范围为 350 至 550°C，损伤速率约为热堆的 10 倍。由于在快堆辐照条件下，不锈钢中的空位形成仅在辐照温度为 400–650°C 时才显著[52]，因此空位肿胀最初不被认为是压水堆（LWR）反应堆芯材料的问题，因为反应堆芯组件的最高温度低于这一温度范围。然而，正如空位肿胀模型[71]所预测的，并且最近由长期压水堆辐照实验[72]所证实，较低的损伤速率可能比高剂量率更具破坏性，因为它们在较低的注量下诱导空位形成，并将空位肿胀范围扩展到较低的温度。 此外，如第 2.1.1 节所述，伽马加热会导致较厚部件的温度升高至 380°C，将其推入奥氏体不锈钢在核电站相关剂量率下容易发生空位肿胀的范围内[52][73]。在压水堆中，挡板-成型螺栓在损伤水平低至 7.5 dpa 时就会产生空位[13]。如果螺栓固定的板也发生肿胀，它们会对螺栓施加应力，从而为辐照辅助应力腐蚀开裂（IASCC）创造条件。迄今为止，已有几例因 IASCC 导致的挡板-成型螺栓开裂的实例[20]。
There are a variety of life extension and power uprate issues for ex-core materials systems used in nuclear power plants. Many of the materials systems, such as piping and heat exchangers, can be considered as replaceable (albeit expensive) components and therefore are not directly influenced by life extension considerations. However, improved predictive knowledge of failure mechanisms can lead to more economical maintenance and replacement schedules, while also improving plant worker and public safety. For the RPV, the primary effect of life extension and power uprate scenarios is an increase in the cumulative neutron fluence to the vessel. As will be discussed in Section 2.2.3, there are significant uncertainties in the long-term embrittlement behavior of irradiated RPV steels due to uncertainties in potentially synergistic, late-emerging nucleation and growth of solute–defect cluster complexes. There are also uncertainties in the maximum useful lifetime of a variety of other non-replaceable (or difficult to replace) materials systems in nuclear power plants. For example, there are nearly 1000 km of power, control and instrumentation cables in a typical nuclear power plant. Considering that electrical cables are known to be susceptible to age-related degradation and shorting (due to moisture, heat, etc.) in applications ranging from electrical appliances to vehicles, homes and factories, which are typically less hostile environments than nuclear power plants, there is interest in developing a suite of non-destructive evaluation techniques to investigate the current performance and expected lifetime of cables [74]. Similarly, the concrete used for safety-related structures in LWRs (e.g. primary containment dome and base slab) is susceptible to a variety of environmental degradation mechanisms that act on the cement matrix and steel reinforcement rods. Research topics of highest interest include long-term degradation mechanisms (>50 years, including ionizing radiation effects), improved noninvasive inspection techniques and improved repair methods [75].
核电站中使用的核芯外材料系统存在多种寿命延长和功率提升问题。许多材料系统，如管道和换热器，可以被视为可替换（尽管昂贵）的部件，因此不受寿命延长考虑的直接影响。然而，对失效机理的预测知识改进可以带来更经济的维护和更换计划，同时提高电站工人和公众的安全。对于反应堆压力容器（RPV），寿命延长和功率提升场景的主要影响是增加容器的累积中子注量。如第 2.2.3 节将讨论的，由于潜在协同作用、晚期出现的溶质-缺陷团簇复合物的成核和生长的不确定性，辐照后的 RPV 钢的长期脆化行为存在显著的不确定性。核电站中各种其他不可替换（或难以替换）材料系统的最大有效寿命也存在不确定性。例如，一个典型的核电站中大约有 1000 公里的电力、控制和仪表电缆。 考虑到电气电缆在从家用电器到车辆、住宅和工厂等应用中容易受到年龄相关退化及短路（由于潮湿、高温等）的影响，而核电站通常环境更为恶劣，因此人们有兴趣开发一套无损评估技术来研究电缆的当前性能和预期寿命[74]。类似地，用于 LWR（轻水反应堆）安全相关结构（例如主包容穹顶和基础板）的混凝土容易受到多种环境退化机制的影响，这些机制作用于水泥基体和钢筋。最感兴趣的研究课题包括长期退化机制（>50 年，包括电离辐射效应）、改进的非侵入式检测技术和改进的修复方法[75]。
核电站中使用的核芯外材料系统存在多种寿命延长和功率提升问题。许多材料系统，如管道和换热器，可以被视为可替换（尽管昂贵）的部件，因此不受寿命延长考虑的直接影响。然而，对失效机理的预测知识改进可以带来更经济的维护和更换计划，同时提高电站工人和公众的安全。对于反应堆压力容器（RPV），寿命延长和功率提升场景的主要影响是增加容器的累积中子注量。如第 2.2.3 节将讨论的，由于潜在协同作用、晚期出现的溶质-缺陷团簇复合物的成核和生长的不确定性，辐照后的 RPV 钢的长期脆化行为存在显著的不确定性。核电站中各种其他不可替换（或难以替换）材料系统的最大有效寿命也存在不确定性。例如，一个典型的核电站中大约有 1000 公里的电力、控制和仪表电缆。 考虑到电气电缆在从家用电器到车辆、住宅和工厂等应用中容易受到年龄相关退化及短路（由于潮湿、高温等）的影响，而核电站通常环境更为恶劣，因此人们有兴趣开发一套无损评估技术来研究电缆的当前性能和预期寿命[74]。类似地，用于 LWR（轻水反应堆）安全相关结构（例如主包容穹顶和基础板）的混凝土容易受到多种环境退化机制的影响，这些机制作用于水泥基体和钢筋。最感兴趣的研究课题包括长期退化机制（>50 年，包括电离辐射效应）、改进的非侵入式检测技术和改进的修复方法[75]。
2.2. Materials degradation challenges in current LWRs
2.2. 当前 LWR 材料退化挑战
Although there are many potential areas of concern for materials in LWRs, the demonstrated highly reliable operating performance of commercial reactors for the past 10–15 years suggests that most of these issues are tractable by appropriate materials selection and engineering design. As described in the following, three specific materials challenges are considered to be of highest importance: (i) exploration of potential further improvements in fuel reliability and operational burn-up limits under normal operating conditions, and safety under transient accident conditions; (ii) corrosion and stress corrosion cracking in reactor components; and (iii) RPV integrity, particularly for life extension scenarios.
尽管在轻水反应堆（LWRs）的材料方面存在许多潜在的担忧，但过去 10-15 年间商业反应堆所展示的高度可靠的运行性能表明，这些问题大多可以通过适当的材料选择和工程设计得到解决。正如下文所述，被认为最重要的三个具体材料挑战是：(i) 在正常运行条件下探索燃料可靠性和运行燃耗极限的进一步潜在改进，以及在瞬态事故条件下的安全性；(ii) 反应堆部件的腐蚀和应力腐蚀开裂；(iii) 压力容器（RPV）完整性，特别是在寿命延长场景下。
尽管在轻水反应堆（LWRs）的材料方面存在许多潜在的担忧，但过去 10-15 年间商业反应堆所展示的高度可靠的运行性能表明，这些问题大多可以通过适当的材料选择和工程设计得到解决。正如下文所述，被认为最重要的三个具体材料挑战是：(i) 在正常运行条件下探索燃料可靠性和运行燃耗极限的进一步潜在改进，以及在瞬态事故条件下的安全性；(ii) 反应堆部件的腐蚀和应力腐蚀开裂；(iii) 压力容器（RPV）完整性，特别是在寿命延长场景下。
2.2.1. Fuel system challenges during normal operations to high burn-up
2.2.1. 正常运行至高燃耗期间的燃料系统挑战
Reliable operation of the fuel assemblies for extended time periods (several cycles, each consisting of 18–24 months of continuous operation) in an extremely harsh radiation environment is of central importance for the economics of nuclear power. Fuel failures (breach of the cladding, allowing escape of some of the gaseous and volatile fission products into the primary coolant, or other damage to the fuel that prevents normal operation) can be managed for a few fuel cladding failures in an operating reactor due to the presence of coolant clean-up systems, but it is undesirable due to increased radiation exposure to personnel, possible leakage of radioactive material, increased inspection and fuel assembly replacement activities during refueling outages, and possible premature shutdown for refueling (if too many fuel failures occur, due to limits on allowable radioactivity in the primary coolant).
在极其恶劣的辐射环境中，燃料组件可靠地运行于较长时间段（数个循环，每个循环包括 18-24 个月的连续运行）对于核能的经济性至关重要。燃料故障（如包壳破裂，允许部分气体和挥发性裂变产物进入一回路冷却剂，或其他损坏燃料导致无法正常运行）由于冷却剂净化系统的存在，在运行中的反应堆中可管理少数燃料包壳故障，但由于增加了人员辐射暴露、放射性物质可能泄漏、换料停堆期间增加的检查和燃料组件更换活动，以及可能因一回路允许放射性物质含量限制而导致的过早换料停堆（如果燃料故障过多），这种情况并不理想。
在极其恶劣的辐射环境中，燃料组件可靠地运行于较长时间段（数个循环，每个循环包括 18-24 个月的连续运行）对于核能的经济性至关重要。燃料故障（如包壳破裂，允许部分气体和挥发性裂变产物进入一回路冷却剂，或其他损坏燃料导致无法正常运行）由于冷却剂净化系统的存在，在运行中的反应堆中可管理少数燃料包壳故障，但由于增加了人员辐射暴露、放射性物质可能泄漏、换料停堆期间增加的检查和燃料组件更换活动，以及可能因一回路允许放射性物质含量限制而导致的过早换料停堆（如果燃料故障过多），这种情况并不理想。
The fuel concept based on monolithic UO2 fuel encased in Zr alloy cladding has undergone remarkable improvements in reliability since it was first developed in the 1950s. The primary advantages associated with Zr for fuel cladding are very low parasitic absorption of neutrons (thereby requiring less initial isotopic enrichment of the fuel to achieve a given burn-up level), and good fabricability and strength; this is countered by relatively poor oxidation behavior in hot water and steam, and anisotropic properties due to the hexagonal close-packed crystal structure. A series of alloying additions have resulted in commercial Zr alloys that have good resistance to oxidation during normal nuclear reactor operations. Alternative cladding options, such as steels, have higher parasitic neutron absorption, and may be susceptible to stress corrosion cracking under certain water chemistry environments in nuclear reactors. Many first-generation LWRs constructed in the 1960s used either austenitic stainless steel (Types 304, 316, 347) or zirconium alloys for the cladding. The stainless steel cladding experienced severe intergranular cracking issues in the BWR coolant environment, due to the highly oxidizing high-temperature steam in combination with radiation hardening that fostered stress corrosion cracking, whereas the Zr alloy cladding was observed to function with adequate reliability [76], [77]. Stainless steel cladding generally exhibited better performance than Zr alloy cladding in first-generation PWRs (fuel pin defect rate of 0.01% vs. 0.1–0.3% for Zr alloy cladding). However, steady improvement of the Zr alloy cladding performance during the 1960s and early 1970s along with the superior (low) parasitic absorption of neutrons by Zr led to nearly universal adoption of Zr alloy cladding by the early 1970s [76], [77]. Cladding performance has continued to improve over the past 30 years, with fleet-averaged fuel pin failure rates of ∼1 × 10−4 in 1980, 2 × 10−5 in 1990 and ∼3 × 10−6 in 2010 [78], [79]. Considering that there are ∼50,000 fuel pins in a LWR, the current fuel reliability means that most reactors now routinely operate without any cladding breaches. This achievement is even more remarkable when one recognizes that the average fuel burn-up more than doubled during the time period of 1980–2010. As a result of this improved fuel performance, along with improvements in other reactor operations such as utilization of online maintenance and improvements in steam generator reliability in PWRs, the capacity factor for fission reactors in the USA has steadily increased from ∼60% in 1980 to ∼90% for the past 10 years. Fig. 10 shows the average fuel burn-up at discharge for US BWR and PWR power plants and the fleet-averaged capacity factors from 1977 to 2010 [78], [79], [80].
基于单晶 UO 2 燃料并包覆在锆合金包壳中的燃料概念，自 20 世纪 50 年代首次开发以来，在可靠性方面取得了显著改进。锆作为燃料包壳材料的主要优势在于其中子寄生吸收率非常低（因此实现给定燃耗水平所需的燃料初始同位素富集量较少），并且具有良好的可加工性和强度；但锆在热水和蒸汽中的氧化行为相对较差，且由于其六方密堆积晶体结构导致存在各向异性。一系列合金化添加剂使得商用锆合金在正常核反应堆运行过程中具有良好的抗氧化性能。其他包壳选项，如不锈钢，具有更高的中子寄生吸收率，并且在某些核反应堆水化学环境下可能易发生应力腐蚀开裂。1960 年代建造的许多第一代轻水反应堆使用了奥氏体不锈钢（304、316、347 型）或锆合金作为包壳材料。 在 BWR 冷却剂环境中，不锈钢包壳出现了严重的沿晶开裂问题，这是由于高温蒸汽的强氧化性与辐射硬化共同作用促进了应力腐蚀开裂。而锆合金包壳则表现出足够的可靠性[76][77]。在第一代压水堆中，不锈钢包壳通常比锆合金包壳表现更好（燃料棒缺陷率为 0.01%对比锆合金包壳的 0.1-0.3%）。然而，20 世纪 60 年代至 70 年代初锆合金包壳性能的稳步提升以及锆对中子（低）寄生吸收的优越性，导致到 20 世纪 70 年代初锆合金包壳几乎得到普遍采用[76][77]。过去 30 年来，包壳性能持续改进，1980 年机组平均燃料棒失效率约为 1×10⁻⁵，1990 年为 2×10⁻⁴，2010 年约为 3×10⁻³[78][79]。考虑到一个轻水堆约有 5 万个燃料棒，当前燃料可靠性意味着大多数反应堆现在常规运行而无包壳破损。 这一成就更加显著，因为人们认识到在 1980 年至 2010 年期间，平均燃料燃耗增加了两倍以上。由于燃料性能的改进，以及其他反应堆操作的改进（如在线维护的利用和压水堆蒸汽发生器可靠性的提高），美国的裂变反应堆容量因子已从 1980 年的约 60%稳步增加到过去 10 年的约 90%。图 10 显示了 1977 年至 2010 年间美国沸水堆和压水堆电站的燃料平均燃耗和机组平均容量因子[78]，[79]，[80]。
基于单晶 UO 2 燃料并包覆在锆合金包壳中的燃料概念，自 20 世纪 50 年代首次开发以来，在可靠性方面取得了显著改进。锆作为燃料包壳材料的主要优势在于其中子寄生吸收率非常低（因此实现给定燃耗水平所需的燃料初始同位素富集量较少），并且具有良好的可加工性和强度；但锆在热水和蒸汽中的氧化行为相对较差，且由于其六方密堆积晶体结构导致存在各向异性。一系列合金化添加剂使得商用锆合金在正常核反应堆运行过程中具有良好的抗氧化性能。其他包壳选项，如不锈钢，具有更高的中子寄生吸收率，并且在某些核反应堆水化学环境下可能易发生应力腐蚀开裂。1960 年代建造的许多第一代轻水反应堆使用了奥氏体不锈钢（304、316、347 型）或锆合金作为包壳材料。 在 BWR 冷却剂环境中，不锈钢包壳出现了严重的沿晶开裂问题，这是由于高温蒸汽的强氧化性与辐射硬化共同作用促进了应力腐蚀开裂。而锆合金包壳则表现出足够的可靠性[76][77]。在第一代压水堆中，不锈钢包壳通常比锆合金包壳表现更好（燃料棒缺陷率为 0.01%对比锆合金包壳的 0.1-0.3%）。然而，20 世纪 60 年代至 70 年代初锆合金包壳性能的稳步提升以及锆对中子（低）寄生吸收的优越性，导致到 20 世纪 70 年代初锆合金包壳几乎得到普遍采用[76][77]。过去 30 年来，包壳性能持续改进，1980 年机组平均燃料棒失效率约为 1×10⁻⁵，1990 年为 2×10⁻⁴，2010 年约为 3×10⁻³[78][79]。考虑到一个轻水堆约有 5 万个燃料棒，当前燃料可靠性意味着大多数反应堆现在常规运行而无包壳破损。 这一成就更加显著，因为人们认识到在 1980 年至 2010 年期间，平均燃料燃耗增加了两倍以上。由于燃料性能的改进，以及其他反应堆操作的改进（如在线维护的利用和压水堆蒸汽发生器可靠性的提高），美国的裂变反应堆容量因子已从 1980 年的约 60%稳步增加到过去 10 年的约 90%。图 10 显示了 1977 年至 2010 年间美国沸水堆和压水堆电站的燃料平均燃耗和机组平均容量因子[78]，[79]，[80]。
Fig. 10. Summary of US LWR burn-up and capacity factors [78], [79], [80].
图 10。美国轻水堆燃耗和容量因子总结[78]，[79]，[80]。
Since a wide range of phenomena can induce fuel failure, a multifaceted approach involving cladding composition, water chemistry and reactor operation modifications has collectively contributed to improved fuel performance [81]. In the 1960s and early 1970s, the relatively poor fuel performance was largely controlled by internal hydriding of the zirconium alloy cladding due to excessive moisture in the ceramic UO2 fuel pellets; this was solved by improved fuel pellet fabrication and use of moisture getters in the fuel plenum in the reactor. In the 1980s and 1990s, as fuel reliability improved, debris fretting of the cladding became a significant problem; debris filters were added to reactors and improved fuel maintenance procedures to minimize introduction of debris were successfully implemented. Numerous water chemistry changes were made in an attempt to balance corrosion issues in different coolant loop components. In the primary circuit of PWRs, Zn was added to suppress steam generator corrosion and stress corrosion cracking and to reduce overall system corrosion, Li was added for pH control (to maintain constant pH as B was added for reactivity control at beginning of reactor cycles) and the pH was increased from ∼6.9 to ∼7.4 in a series of steps to control CRUD deposits on the surfaces of the fuel cladding [79]. In BWRs, noble metal additions and H were added to reduce stress corrosion cracking. As burn-ups increased, pellet–cladding interactions became more prominent due to increased fuel swelling. The closure of the initial gap between the fuel pellets and the cladding can produce a variety of mechanical stress effects as well as multiple interactions with aggressive fission products. One successful approach has been to utilize multiple-layered cladding (e.g. utilizing a soft Zr liner as a compliant barrier liner between the fuel and cladding) [81], [82]. The fuel pin diameter and cladding thickness were also decreased to reduce heat flux and facilitate high burn-up (along with operational changes that limited the reactor ramp rate during start-up) [81], with current pellet diameters of ∼10 mm and cladding thickness of ∼0.6 mm for BWRs and PWRs [83]. The current major fuel cladding degradation phenomena for high burn-ups [79], [84] include grid-to-rod fretting in PWRs and cladding corrosion/oxidation/CRUD deposition and pellet-cladding interactions in both BWRs and PWRs [85]. Grid-to-rod fretting is induced by coolant-flow-induced vibration of the fuel cladding against the horizontal grid assembly spacer. At high burn-ups the spring force between the cladding and grid assembly is relaxed due to irradiation creep [12]. Grid-to-rod fretting can be reduced by providing additional stiffening of the fuel assembly and by design changes to minimize fuel rod or assembly vibration, as well as to suppress cross flow and jetting (grid-to-rod fretting is not a major issue in BWRs due to the presence of channel boxes that restrict flow between fuel assemblies).
由于多种现象都可能引发燃料失效，因此涉及包壳成分、水化学和反应堆运行改进的多方面方法共同促成了燃料性能的提升[81]。在 20 世纪 60 年代和 70 年代初，由于陶瓷 UO₂燃料颗粒中水分过多导致锆合金包壳内部发生氢化，燃料性能相对较差的问题主要受此影响；这一问题通过改进燃料颗粒制造工艺和在反应堆燃料腔中使用吸湿剂得到解决。在 20 世纪 80 年代和 90 年代，随着燃料可靠性的提高，包壳的碎片磨损成为了一个显著问题；反应堆中增加了碎片过滤器，并成功实施了改进的燃料维护程序以最大限度减少碎片的引入。为了平衡不同冷却剂回路组件的腐蚀问题，进行了多次水化学方面的调整。 在压水堆（PWR）的一回路中，添加了 Zn 以抑制蒸汽发生器腐蚀和应力腐蚀开裂，并减少整个系统的腐蚀；添加了 Li 以控制 pH 值（在反应堆循环开始时添加 B 以控制反应性，以保持 pH 值恒定），并将 pH 值从约 6.9 逐步提高到约 7.4，以控制燃料包壳表面的 CRUD 沉积物[79]。在沸水堆（BWR）中，添加了贵金属和 H 以减少应力腐蚀开裂。随着燃耗的增加，由于燃料肿胀加剧，芯块-包壳相互作用变得更加突出。燃料芯块与包壳之间的初始间隙闭合会产生多种机械应力效应以及与活泼裂变产物的多重相互作用。一种成功的方法是采用多层包壳（例如，利用软质 Zr 内衬作为燃料与包壳之间的柔性屏障内衬）[81][82]。 燃料棒直径和包壳厚度也减小了，以降低热通量并促进高燃耗（同时还包括限制启动期间反应堆斜坡速率的操作变更）[81]，目前沸水堆（BWR）和压水堆（PWR）的燃料丸直径约为 10 毫米，包壳厚度约为 0.6 毫米[83]。目前高燃耗的主要燃料包壳劣化现象[79][84]包括压水堆中的棒栅磨损，以及沸水堆和压水堆中的包壳腐蚀/氧化/CRUD 沉积和燃料丸-包壳相互作用[85]。棒栅磨损是由冷却剂流动引起的燃料包壳对水平栅格组件的振动所诱导的。在高燃耗下，由于辐照蠕变[12]，包壳与栅格组件之间的弹簧力会减弱。通过提供燃料组件的额外加固，以及通过设计变更来最小化燃料棒或组件的振动，并抑制横向流动和喷射（由于通道箱的存在限制了燃料组件之间的流动，棒栅磨损在沸水堆中不是主要问题）。
由于多种现象都可能引发燃料失效，因此涉及包壳成分、水化学和反应堆运行改进的多方面方法共同促成了燃料性能的提升[81]。在 20 世纪 60 年代和 70 年代初，由于陶瓷 UO₂燃料颗粒中水分过多导致锆合金包壳内部发生氢化，燃料性能相对较差的问题主要受此影响；这一问题通过改进燃料颗粒制造工艺和在反应堆燃料腔中使用吸湿剂得到解决。在 20 世纪 80 年代和 90 年代，随着燃料可靠性的提高，包壳的碎片磨损成为了一个显著问题；反应堆中增加了碎片过滤器，并成功实施了改进的燃料维护程序以最大限度减少碎片的引入。为了平衡不同冷却剂回路组件的腐蚀问题，进行了多次水化学方面的调整。 在压水堆（PWR）的一回路中，添加了 Zn 以抑制蒸汽发生器腐蚀和应力腐蚀开裂，并减少整个系统的腐蚀；添加了 Li 以控制 pH 值（在反应堆循环开始时添加 B 以控制反应性，以保持 pH 值恒定），并将 pH 值从约 6.9 逐步提高到约 7.4，以控制燃料包壳表面的 CRUD 沉积物[79]。在沸水堆（BWR）中，添加了贵金属和 H 以减少应力腐蚀开裂。随着燃耗的增加，由于燃料肿胀加剧，芯块-包壳相互作用变得更加突出。燃料芯块与包壳之间的初始间隙闭合会产生多种机械应力效应以及与活泼裂变产物的多重相互作用。一种成功的方法是采用多层包壳（例如，利用软质 Zr 内衬作为燃料与包壳之间的柔性屏障内衬）[81][82]。 燃料棒直径和包壳厚度也减小了，以降低热通量并促进高燃耗（同时还包括限制启动期间反应堆斜坡速率的操作变更）[81]，目前沸水堆（BWR）和压水堆（PWR）的燃料丸直径约为 10 毫米，包壳厚度约为 0.6 毫米[83]。目前高燃耗的主要燃料包壳劣化现象[79][84]包括压水堆中的棒栅磨损，以及沸水堆和压水堆中的包壳腐蚀/氧化/CRUD 沉积和燃料丸-包壳相互作用[85]。棒栅磨损是由冷却剂流动引起的燃料包壳对水平栅格组件的振动所诱导的。在高燃耗下，由于辐照蠕变[12]，包壳与栅格组件之间的弹簧力会减弱。通过提供燃料组件的额外加固，以及通过设计变更来最小化燃料棒或组件的振动，并抑制横向流动和喷射（由于通道箱的存在限制了燃料组件之间的流动，棒栅磨损在沸水堆中不是主要问题）。
Looking to the future, there is interest in continuing to increase the fuel burn-up values while maintaining or improving the reactor capacity factor and continuing to reduce the fuel failure rate toward zero (individual fuel pin failure probability of <1 × 10−7). Decreased overall fuel cycle costs, improved operational flexibility and reduced volume of radioactive waste are all benefits associated with increasing burn-up [81]. Grid-to-rod fretting challenges may be more pronounced in reactors approved for power uprates due to increased coolant flow rates [85], and higher burn-ups would produce additional irradiation creep relaxation of the springs in the grid assembly, which could exacerbate the fretting issue. Advanced thermal hydraulic modeling and other analyses would be useful to determine if new designs for the spacer grid might reduce the magnitude of the fretting. Distortions in the fuel bundles, PWR control rod guide tubes and BWR channel plates at high burn-ups (including bowing of constrained components) due to anisotropic irradiation growth of the zirconium alloy could impede the motion of control rods or blades and therefore need to be carefully evaluated. Finally, oxidation of the cladding at very high burn-ups may require the development of a new generation of ultra-high oxidation-resistant claddings. Oxidation introduces two potential degradation mechanisms: loss of sound metal for the structural function of the cladding and the possible introduction into the cladding of hydrogen formed as a result of the reduction of steam during the oxidation process (including the formation of hydride precipitates, which may serve as internal stress concentrators). An additional concern associated with hydrogen pick-up is the potential for reorientation of the hydride precipitates from circumferential to radial direction as a result of spent fuel handling procedures after the fuel is withdrawn from the reactor [86]. This reorientation of the hydride precipitates could lead to cladding failure in used fuel during handling and transport, with a consequential effect on the cost for safe management of used fuel. Current international safety regulations for loss of coolant scenarios specify that the oxidized cladding should be less than 15–17% of the wall thickness, which corresponds to an oxide thickness of about 100 μm. This limits historical cladding materials such as Zircaloy-4 to burn-up levels below ∼50–60 GWd MTU−1, whereas more recent oxidation-resistant Zr alloys, such as optimized ZIRLO and M5/AXIOM, appear to exhibit suitable in-pile oxidation behavior for burn-ups much higher than 70 GWd MTU−1 [87], [88], [89]. The slightly higher temperatures associated with power uprate activities should pose an additional challenge, since the oxidation rate will be correspondingly increased due to the higher temperature.
展望未来，人们希望继续提高燃料燃耗值，同时保持或提高反应堆容量因子，并持续降低燃料失效率直至零（单个燃料棒失效概率小于 1×10^-6）。降低整体燃料循环成本、提高运行灵活性和减少放射性废物量都是提高燃耗带来的好处[81]。由于冷却剂流量增加，在批准功率提升的反应堆中，棒栅间松动问题可能更为突出[85]，而更高的燃耗会产生栅格组件中弹簧的额外辐照蠕变弛豫，这可能加剧松动问题。先进的传热水力模型和其他分析将有助于确定新的栅格设计是否可能降低松动程度。由于锆合金各向异性辐照生长，在高燃耗下燃料组件、压水堆控制棒导向管和沸水堆通道板可能发生扭曲（包括受约束部件的弯曲），这可能会阻碍控制棒或叶片的运动，因此需要仔细评估。 最后，在非常高的燃耗下，包壳的氧化可能需要开发新一代的超高抗氧化性包壳。氧化引入了两种潜在的性能退化机制：一是包壳的结构功能因金属损失而减弱，二是可能因氧化过程中蒸汽的还原反应而在包壳中引入氢（包括形成氢化物沉淀，这些沉淀可能成为内部应力集中点）。与氢吸收相关的一个额外问题是，在燃料从反应堆中取出后的乏燃料处理过程中，氢化物沉淀可能从周向重新定向为径向方向[86]。这种氢化物沉淀的重新定向可能导致乏燃料在处理和运输过程中包壳失效，进而影响乏燃料安全管理的成本。当前国际安全规范针对失水事故场景规定，氧化后的包壳厚度应小于壁厚的 15-17%，这相当于氧化层厚度约为 100 微米。 这限制了历史上的包壳材料如 Zircaloy-4 的使用，使其燃耗水平低于~50–60 GWd MTU −1 ，而较新的抗氧化 Zr 合金，如优化的 ZIRLO 和 M5/AXIOM，似乎在堆内表现出适合于燃耗远高于 70 GWd MTU −1 的氧化行为[87], [88], [89]。与功率提升活动相关的稍高温度应该会构成额外的挑战，因为氧化速率会因温度更高而相应增加。
展望未来，人们希望继续提高燃料燃耗值，同时保持或提高反应堆容量因子，并持续降低燃料失效率直至零（单个燃料棒失效概率小于 1×10^-6）。降低整体燃料循环成本、提高运行灵活性和减少放射性废物量都是提高燃耗带来的好处[81]。由于冷却剂流量增加，在批准功率提升的反应堆中，棒栅间松动问题可能更为突出[85]，而更高的燃耗会产生栅格组件中弹簧的额外辐照蠕变弛豫，这可能加剧松动问题。先进的传热水力模型和其他分析将有助于确定新的栅格设计是否可能降低松动程度。由于锆合金各向异性辐照生长，在高燃耗下燃料组件、压水堆控制棒导向管和沸水堆通道板可能发生扭曲（包括受约束部件的弯曲），这可能会阻碍控制棒或叶片的运动，因此需要仔细评估。 最后，在非常高的燃耗下，包壳的氧化可能需要开发新一代的超高抗氧化性包壳。氧化引入了两种潜在的性能退化机制：一是包壳的结构功能因金属损失而减弱，二是可能因氧化过程中蒸汽的还原反应而在包壳中引入氢（包括形成氢化物沉淀，这些沉淀可能成为内部应力集中点）。与氢吸收相关的一个额外问题是，在燃料从反应堆中取出后的乏燃料处理过程中，氢化物沉淀可能从周向重新定向为径向方向[86]。这种氢化物沉淀的重新定向可能导致乏燃料在处理和运输过程中包壳失效，进而影响乏燃料安全管理的成本。当前国际安全规范针对失水事故场景规定，氧化后的包壳厚度应小于壁厚的 15-17%，这相当于氧化层厚度约为 100 微米。 这限制了历史上的包壳材料如 Zircaloy-4 的使用，使其燃耗水平低于~50–60 GWd MTU −1 ，而较新的抗氧化 Zr 合金，如优化的 ZIRLO 和 M5/AXIOM，似乎在堆内表现出适合于燃耗远高于 70 GWd MTU −1 的氧化行为[87], [88], [89]。与功率提升活动相关的稍高温度应该会构成额外的挑战，因为氧化速率会因温度更高而相应增加。
The fuel pellet experiences the most severe temperatures, radiation damage and chemical transmutation environment of any component in a nuclear reactor. The cumulative amount of radiation damage approaches 1000 dpa for a typical fuel pellet, with 5% or more of the original uranium atoms converted into ∼10 at.% or more fission products [90]. The microstructure of the initially monolithic polycrystalline UO2 fuel undergoes dramatic changes during high burn-up irradiation due to the intense displacement damage and chemical transmutations, along with high power densities and the relatively low thermal conductivity of the UO2 fuel that produce temperature gradients >1000°C cm−1. Of particular interest is the formation at burn-up levels above ∼50 GWd MTU−1 of nanoscale polygonized grains in the outer “rim” region of the fuel pellet, where the burn-up levels are highest and the temperature is lowest [90], [91], [92]. There was initially concern that this new high burn-up structure might exhibit inferior fission gas retention, thermal conductivity or other poor fuel properties. However, recent results indicate that the high burn-up structure generally exhibits behavior comparable or favorable to the lower burn-up microstructure [90]. Good fuel behavior has been observed in test irradiations for fuel irradiations up to 83 GWd MTU−1 [93]. Exploratory research is also being performed on a variety of alternative fuel forms that are a significant departure from monolithic sintered UO2, including inert matrix fuels [94], metallic or ceramic matrix microencapsulated article fuels [95], and nanocrystalline oxide fuels [96], which may offer some advantages in achieving very higher burn-ups.
燃料芯块是核反应堆中承受最严酷温度、辐射损伤和化学转化的部件。对于典型的燃料芯块，累积的辐射损伤量接近 1000 dpa，其中 5%或更多的原始铀原子转化为约 10 at.%或更多的裂变产物[90]。最初为单相多晶 UO 2 燃料的微观结构在高燃耗辐照期间由于强烈的位移损伤和化学转化，加上 UO 2 燃料的高功率密度和相对较低的热导率，导致温度梯度超过 1000°C cm −1 ，发生了剧烈变化。特别值得关注的是，在燃耗水平超过约 50 GWd MTU −1 时，在燃料芯块外部的“边缘”区域（燃耗水平最高、温度最低处）形成了纳米级多边形晶粒[90]、[91]、[92]。最初人们担心这种新的高燃耗结构可能会表现出较差的裂变气体保持能力、热导率或其他不良燃料特性。 然而，最近的研究结果表明，高燃耗结构通常表现出与低燃耗微观结构相当或更有利的特性[90]。在高达 83 GWd MTU −1 的燃料辐照试验中观察到了良好的燃料行为[93]。同时也在进行多种替代燃料形式的开创性研究，这些形式与单晶烧结 UO 2 有显著区别，包括惰性基质燃料[94]、金属或陶瓷基质微胶囊化颗粒燃料[95]以及纳米晶氧化物燃料[96]，这些燃料在实现非常高的燃耗方面可能具有一些优势。
燃料芯块是核反应堆中承受最严酷温度、辐射损伤和化学转化的部件。对于典型的燃料芯块，累积的辐射损伤量接近 1000 dpa，其中 5%或更多的原始铀原子转化为约 10 at.%或更多的裂变产物[90]。最初为单相多晶 UO 2 燃料的微观结构在高燃耗辐照期间由于强烈的位移损伤和化学转化，加上 UO 2 燃料的高功率密度和相对较低的热导率，导致温度梯度超过 1000°C cm −1 ，发生了剧烈变化。特别值得关注的是，在燃耗水平超过约 50 GWd MTU −1 时，在燃料芯块外部的“边缘”区域（燃耗水平最高、温度最低处）形成了纳米级多边形晶粒[90]、[91]、[92]。最初人们担心这种新的高燃耗结构可能会表现出较差的裂变气体保持能力、热导率或其他不良燃料特性。 然而，最近的研究结果表明，高燃耗结构通常表现出与低燃耗微观结构相当或更有利的特性[90]。在高达 83 GWd MTU −1 的燃料辐照试验中观察到了良好的燃料行为[93]。同时也在进行多种替代燃料形式的开创性研究，这些形式与单晶烧结 UO 2 有显著区别，包括惰性基质燃料[94]、金属或陶瓷基质微胶囊化颗粒燃料[95]以及纳米晶氧化物燃料[96]，这些燃料在实现非常高的燃耗方面可能具有一些优势。
2.2.2. Corrosion and stress corrosion cracking (SCC) in structural materials
2.2.2. 结构材料中的腐蚀和应力腐蚀开裂（SCC）
The temperature range of the water coolant in LWRs is generally between 275 and 325 °C, and spans the saturated (BWR) to sub-cooled (PWR) regimes. Although LWRs have been in operation for over 50 years, corrosion remains a significant concern that will become even more important with age. Corrosion occurs in all of the major systems exposed to a water environment, including the reactor core, steam generator, turbine, condenser and piping, valves and fittings, and in a wide variety of alloys such as carbon and low alloy steels used in piping and turbine components, stainless steel used in core internals and primary flow circuits and the condenser, nickel-base alloys in the steam generator and in reactor vessel penetrations and welds, and zirconium alloy fuel cladding [17], [19], [20], [97].
轻水反应堆（LWRs）中冷却水的温度范围通常在 275 至 325°C 之间，涵盖了饱和（BWR）至过冷（PWR）的各个阶段。尽管轻水反应堆已运行超过 50 年，但腐蚀仍然是一个重要问题，并且随着时间推移会变得更加突出。腐蚀发生在所有暴露于水环境的主要系统中，包括反应堆堆芯、蒸汽发生器、汽轮机、冷凝器、管道、阀门和管件，以及用于管道和汽轮机部件的碳钢和低合金钢、用于堆芯内部和一回路以及冷凝器的不锈钢、用于蒸汽发生器和反应堆容器贯穿件及焊缝的镍基合金，以及锆合金燃料包壳[17]、[19]、[20]、[97]。
轻水反应堆（LWRs）中冷却水的温度范围通常在 275 至 325°C 之间，涵盖了饱和（BWR）至过冷（PWR）的各个阶段。尽管轻水反应堆已运行超过 50 年，但腐蚀仍然是一个重要问题，并且随着时间推移会变得更加突出。腐蚀发生在所有暴露于水环境的主要系统中，包括反应堆堆芯、蒸汽发生器、汽轮机、冷凝器、管道、阀门和管件，以及用于管道和汽轮机部件的碳钢和低合金钢、用于堆芯内部和一回路以及冷凝器的不锈钢、用于蒸汽发生器和反应堆容器贯穿件及焊缝的镍基合金，以及锆合金燃料包壳[17]、[19]、[20]、[97]。
Early corrosion problems stemmed from “epidemics” that were generally precipitated by improper water chemistry control; ingress of chlorides that induced pitting in steam turbine discs and blades, and pitting and stress corrosion cracking in stainless steel, or poor secondary side pH control that resulted in wastage and crevice corrosion in steam generator tubes, or poor microstructure or alloy chemistry control; high corrosion rates of zirconium fuel cladding, stress corrosion cracking of stainless steel BWR piping due to sensitization or weld knife-line attack [17]. More recently, corrosion degradation has emerged in the forms of stress corrosion cracking of stainless steel steam lines and nickel-base steam generator tubes and reactor vessel penetrations, flow assisted corrosion in low alloy steels, and nodular corrosion, shadow corrosion, CRUD-induced localized corrosion and fretting of zirconium alloy fuel cladding [18]. Still more recently, irradiation has emerged to play an increasingly important role in irradiation-assisted stress corrosion cracking (IASCC) and irradiation-accelerated corrosion [21], [22]. As plants age, the most important corrosion issues will center about stress corrosion cracking, and the accelerating role of irradiation in both IASCC and corrosion. A recent example of corrosion-induced degradation occurred in 2002 at the Davis-Besse reactor (PWR), when stress corrosion cracking of the weld metal between the vessel head and a control rod drive housing on the vessel head caused coolant to leak onto the head. Evaporation of water on the hot surface resulted in a concentrated boric acid solution that resulted in corrosion to such an extent as to create a hole in the head nearly the size of a soccer ball. The ∼10 mm thick stainless steel weld metal liner prevented a loss-of-coolant accident.
早期的腐蚀问题源于“流行病”，这些通常由不适当的水化学控制引发；氯离子侵入导致汽轮机盘和叶片出现点蚀，不锈钢出现点蚀和应力腐蚀开裂，或二次侧 pH 控制不良导致蒸汽发生器管路出现浪费和缝隙腐蚀，或微观结构或合金化学控制不佳；锆燃料包壳的高腐蚀速率，或由于敏化或焊缝刀边侵蚀导致的不锈钢沸水反应堆管道的应力腐蚀开裂[17]。最近，腐蚀退化以不锈钢蒸汽管道和镍基蒸汽发生器管路及反应堆容器穿透的应力腐蚀开裂、低合金钢的流动辅助腐蚀、锆合金燃料包壳的球状腐蚀、阴影腐蚀、CRUD 诱导的局部腐蚀和磨损等形式出现[18]。更近一些时候，辐照被发现越来越多地发挥重要作用，在辐照辅助应力腐蚀开裂（IASCC）和辐照加速腐蚀中[21][22]。 随着植物老化，最重要的腐蚀问题将集中在应力腐蚀开裂，以及辐照在应力腐蚀开裂和腐蚀中的加速作用。一个最近因腐蚀引起的退化案例发生在 2002 年的戴维斯-贝斯反应堆（压水堆），当容器盖和控制杆驱动 housing 之间的焊缝金属发生应力腐蚀开裂时，冷却剂泄漏到容器盖上。热表面上的水分蒸发导致形成浓缩的硼酸溶液，腐蚀程度如此严重，以至于在容器盖上形成了一个几乎有足球那么大的孔洞。厚约 10 毫米的不锈钢焊缝金属内衬防止了失水事故。
早期的腐蚀问题源于“流行病”，这些通常由不适当的水化学控制引发；氯离子侵入导致汽轮机盘和叶片出现点蚀，不锈钢出现点蚀和应力腐蚀开裂，或二次侧 pH 控制不良导致蒸汽发生器管路出现浪费和缝隙腐蚀，或微观结构或合金化学控制不佳；锆燃料包壳的高腐蚀速率，或由于敏化或焊缝刀边侵蚀导致的不锈钢沸水反应堆管道的应力腐蚀开裂[17]。最近，腐蚀退化以不锈钢蒸汽管道和镍基蒸汽发生器管路及反应堆容器穿透的应力腐蚀开裂、低合金钢的流动辅助腐蚀、锆合金燃料包壳的球状腐蚀、阴影腐蚀、CRUD 诱导的局部腐蚀和磨损等形式出现[18]。更近一些时候，辐照被发现越来越多地发挥重要作用，在辐照辅助应力腐蚀开裂（IASCC）和辐照加速腐蚀中[21][22]。 随着植物老化，最重要的腐蚀问题将集中在应力腐蚀开裂，以及辐照在应力腐蚀开裂和腐蚀中的加速作用。一个最近因腐蚀引起的退化案例发生在 2002 年的戴维斯-贝斯反应堆（压水堆），当容器盖和控制杆驱动 housing 之间的焊缝金属发生应力腐蚀开裂时，冷却剂泄漏到容器盖上。热表面上的水分蒸发导致形成浓缩的硼酸溶液，腐蚀程度如此严重，以至于在容器盖上形成了一个几乎有足球那么大的孔洞。厚约 10 毫米的不锈钢焊缝金属内衬防止了失水事故。
2.2.2.1. Stress corrosion cracking
2.2.2.1. 应力腐蚀开裂
Today, the major stress corrosion cracking issues are in the nickel-base alloys used in steam generators and in vessel penetrations and stainless steels used in the reactor core. Since steam generator tubes comprise some 75% of the surface area of the primary circuit in contact with the coolant, their performance is critical to that of the reactor. Fig. 11 shows that, while many degradation modes of alloy 600 steam generator tubes exist, SCC has been the dominant failure mode over the past 25 years [98]. SCC of nickel-base alloys is very sensitive to composition and water chemistry. Fig. 12 shows the propensity for cracking of austenitic alloys as a function of nickel content in both pure water and 0.1% NaCl; this cracking can either be transgranular (TGSCC) or intergranular (IGSCC) [10]. Note that SCC in chloride is at a maximum at the extremes in nickel content, but SCC in pure water occurs only at high nickel concentrations. Unfortunately, alloy 600 used originally in steam generators and vessel penetrations contains approximately 78% Ni and therefore is highly susceptible to SCC even in pure water. Neither microstructure modification (thermal treatment) nor water chemistry control has been able to mitigate SCC in alloy 600. Staehle and Gorman identified seven different SCC modes defined by their potential–pH combinations [97]. In fact, SCC in alloy 600 appears to track the Ni–NiO stability line, in which cracking is at a maximum at the line and drops off both above and below it (Fig. 13) [10]. Since this range of potential and pH spans those typically achievable in service, control of SCC in alloy 600 has proved very difficult. As a result, alloy 600 components have been replaced by alloy 690, which has a nominal composition of 60Ni–30Cr–10Fe and exhibits much higher resistance to SCC in pure water, chlorides and alkaline solutions. However, historically, alloy 600 was found to be difficult to crack in pure high-temperature water in the laboratory, prompting the belief that it would be resistant in service. In fact, even after years in operation, few incidents of cracking were reported. Over time, the incubation period was discovered to be of the order of 11–12 years, and eventually, all plants began to experience cracking. This has led to widespread replacement by steam generators constructed from alloy 690. The concern today is that the superior resistance of alloy 690 to SCC may be due to a longer incubation period and that, in fact, it may eventually begin to crack. Current efforts are focused on trying to determine if there are microstructures, processing routes or water chemistries through which alloy 690 is susceptible to SCC. Recently, it was found that a single cold rolling operation to 20–30% reduction in thickness increases the crack growth rate in pure water by over a factor of 100 [99]. Similar research activities are in progress to identify susceptibility to crack initiation.
如今，主要的应力腐蚀开裂问题出现在用于蒸汽发生器和容器贯穿件的镍基合金以及反应堆堆芯使用的奥氏体不锈钢上。由于蒸汽发生器管子占冷却剂接触的一回路表面积的约 75%，因此它们的工作性能对反应堆至关重要。图 11 表明，虽然合金 600 蒸汽发生器管子存在多种退化模式，但在过去 25 年中，应力腐蚀开裂一直是主要的失效模式[98]。镍基合金的应力腐蚀开裂对成分和水化学非常敏感。图 12 显示了奥氏体合金在纯水和 0.1% NaCl 中的开裂倾向，作为镍含量的函数；这种开裂可以是沿晶（TGSCC）或穿晶（IGSCC）[10]。请注意，在氯化物中，应力腐蚀开裂在镍含量极端时达到最大值，但在纯水中仅在镍浓度高时发生。不幸的是，最初用于蒸汽发生器和容器贯穿件的合金 600 含有约 78%的 Ni，因此在纯水中也非常容易发生应力腐蚀开裂。 微结构改性（热处理）和水化学控制均未能缓解合金 600 的应力腐蚀开裂（SCC）。斯塔赫尔和戈尔曼根据其电位-pH 组合定义了七种不同的 SCC 模式[97]。事实上，合金 600 中的 SCC 似乎遵循 Ni-NiO 稳定性曲线，在该曲线上裂纹最多，曲线上下裂纹数量均减少（图 13）[10]。由于这个电位和 pH 范围涵盖了通常在服役中可达到的范围，因此控制合金 600 的 SCC 被证明非常困难。结果，合金 600 部件已被合金 690 取代，其名义成分为 60Ni-30Cr-10Fe，在纯水、氯化物和碱性溶液中表现出更高的抗 SCC 性能。然而，历史上在实验室中纯高温水中发现合金 600 难以开裂，从而促成了其在服役中具有抗性的信念。事实上，即使运行多年后，也鲜有开裂事件报道。随着时间的推移，发现潜伏期约为 11-12 年，最终所有工厂都开始出现开裂。 这导致了广泛使用由 690 合金制成的蒸汽发生器。如今的问题是，690 合金对应力腐蚀开裂（SCC）的优越抗性可能是因为潜伏期更长，实际上它最终可能会开始开裂。当前的研究重点是尝试确定是否存在微观结构、加工路线或水化学条件，使得 690 合金易受 SCC 影响。最近发现，单次冷轧使厚度减少 20-30%会增加纯水中裂纹扩展速率超过 100 倍[99]。类似的科研活动正在进行中，以确定对裂纹萌生的敏感性。
如今，主要的应力腐蚀开裂问题出现在用于蒸汽发生器和容器贯穿件的镍基合金以及反应堆堆芯使用的奥氏体不锈钢上。由于蒸汽发生器管子占冷却剂接触的一回路表面积的约 75%，因此它们的工作性能对反应堆至关重要。图 11 表明，虽然合金 600 蒸汽发生器管子存在多种退化模式，但在过去 25 年中，应力腐蚀开裂一直是主要的失效模式[98]。镍基合金的应力腐蚀开裂对成分和水化学非常敏感。图 12 显示了奥氏体合金在纯水和 0.1% NaCl 中的开裂倾向，作为镍含量的函数；这种开裂可以是沿晶（TGSCC）或穿晶（IGSCC）[10]。请注意，在氯化物中，应力腐蚀开裂在镍含量极端时达到最大值，但在纯水中仅在镍浓度高时发生。不幸的是，最初用于蒸汽发生器和容器贯穿件的合金 600 含有约 78%的 Ni，因此在纯水中也非常容易发生应力腐蚀开裂。 微结构改性（热处理）和水化学控制均未能缓解合金 600 的应力腐蚀开裂（SCC）。斯塔赫尔和戈尔曼根据其电位-pH 组合定义了七种不同的 SCC 模式[97]。事实上，合金 600 中的 SCC 似乎遵循 Ni-NiO 稳定性曲线，在该曲线上裂纹最多，曲线上下裂纹数量均减少（图 13）[10]。由于这个电位和 pH 范围涵盖了通常在服役中可达到的范围，因此控制合金 600 的 SCC 被证明非常困难。结果，合金 600 部件已被合金 690 取代，其名义成分为 60Ni-30Cr-10Fe，在纯水、氯化物和碱性溶液中表现出更高的抗 SCC 性能。然而，历史上在实验室中纯高温水中发现合金 600 难以开裂，从而促成了其在服役中具有抗性的信念。事实上，即使运行多年后，也鲜有开裂事件报道。随着时间的推移，发现潜伏期约为 11-12 年，最终所有工厂都开始出现开裂。 这导致了广泛使用由 690 合金制成的蒸汽发生器。如今的问题是，690 合金对应力腐蚀开裂（SCC）的优越抗性可能是因为潜伏期更长，实际上它最终可能会开始开裂。当前的研究重点是尝试确定是否存在微观结构、加工路线或水化学条件，使得 690 合金易受 SCC 影响。最近发现，单次冷轧使厚度减少 20-30%会增加纯水中裂纹扩展速率超过 100 倍[99]。类似的科研活动正在进行中，以确定对裂纹萌生的敏感性。
Fig. 11. Failure modes of mill-annealed Alloy 600 steam generator tubes in US PWRs over a 38-year period [98].
图 11. 美国压水堆（US PWR）中经轧制退火处理的 Alloy 600 蒸汽发生器管材在 38 年内的失效模式[98]。
Fig. 12. SCC severity of austenitic alloys as a function of nickel content in pure water and 0.1% sodium chloride (courtesy of R.W. Staehle [10]).
图 12.奥氏体合金的应力腐蚀开裂严重程度与镍含量的关系（纯水和 0.1%氯化钠环境）（致谢 R.W. Staehle [10]）。
Fig. 13. Modes of SCC in Alloy 600 affected by environmental chemistry (courtesy of R.W. Staehle [10]). Regimes in the figure are: AcSCC: acidic-induced SCC; AkSCC: alkaline-induced SCC; HPSCC: high potential-induced SCC; LPSCC: low potential-induced SCC; AkIGC: alkaline-induced intergranular corrosion; PbSCC: lead-induced SCC and Sy-SCC: sulfide-induced SCC.
图 13.环境化学对合金 600 应力腐蚀开裂模式的影响（致谢 R.W. Staehle [10]）。图中所示模式包括：AcSCC：酸性诱导的应力腐蚀开裂；AkSCC：碱性诱导的应力腐蚀开裂；HPSCC：高电位诱导的应力腐蚀开裂；LPSCC：低电位诱导的应力腐蚀开裂；AkIGC：碱性诱导的晶间腐蚀；PbSCC：铅诱导的应力腐蚀开裂；S y- SCC：硫化物诱导的应力腐蚀开裂。
2.2.2.2. Irradiation-assisted stress corrosion cracking
2.2.2.2. 辐照辅助应力腐蚀开裂
IASCC of austenitic stainless steels and some nickel-base alloys has presented a significant problem in ensuring the integrity of LWR core components. IASCC is a generic challenge as it apparently cuts across all LWR designs and materials. Table 4 shows that IASCC has been observed in at least four water reactor designs, 11 different alloys and dozens of components. The specific effects of irradiation on IASCC are classified into two categories: water chemistry and microstructure [21], [22]. Water chemistry effects include radiolysis and its effects on corrosion potential, and the effects of corrosion potential on IASCC. Microstructure effects include radiation-induced segregation, irradiated microstructure, swelling and creep, and H and He generation. Leading mechanisms proposed to explain the roles of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and radiation-induced solute segregation of minor elements [21], [22].
奥氏体不锈钢和某些镍基合金的辐照脆化（IASCC）在确保轻水反应堆（LWR）核心部件完整性方面构成了一个重要问题。IASCC 是一个普遍挑战，因为它显然跨越了所有 LWR 设计和材料。表 4 表明，IASCC 已在至少四种水反应堆设计中、11 种不同合金和数十个部件中被观察到。辐照对 IASCC 的具体影响可分为两类：水化学和微观结构[21], [22]。水化学效应包括辐射分解及其对腐蚀电位的影响，以及腐蚀电位对 IASCC 的影响。微观结构效应包括辐射诱导偏析、辐照后的微观结构、肿胀和蠕变，以及 H 和 He 的产生。为解释辐射在应力腐蚀开裂（SCC）过程中的作用，提出的首要机制包括：辐射分解、裂纹尖端应变率、晶界铬贫化、辐照硬化、局部变形和微量元素的辐射诱导溶剂偏析[21], [22]。
奥氏体不锈钢和某些镍基合金的辐照脆化（IASCC）在确保轻水反应堆（LWR）核心部件完整性方面构成了一个重要问题。IASCC 是一个普遍挑战，因为它显然跨越了所有 LWR 设计和材料。表 4 表明，IASCC 已在至少四种水反应堆设计中、11 种不同合金和数十个部件中被观察到。辐照对 IASCC 的具体影响可分为两类：水化学和微观结构[21], [22]。水化学效应包括辐射分解及其对腐蚀电位的影响，以及腐蚀电位对 IASCC 的影响。微观结构效应包括辐射诱导偏析、辐照后的微观结构、肿胀和蠕变，以及 H 和 He 的产生。为解释辐射在应力腐蚀开裂（SCC）过程中的作用，提出的首要机制包括：辐射分解、裂纹尖端应变率、晶界铬贫化、辐照硬化、局部变形和微量元素的辐射诱导溶剂偏析[21], [22]。
Table 4. Summary of observed IASCC issues in LWR components [21], [22].
表 4. LWR 部件中观察到的 IASCC 问题总结[21], [22].
| Component 部件 | Material 材料 | Reactor type 反应堆类型 | Possible sources of stress 可能的应力来源 |
|---|---|---|---|
| Fuel cladding 燃料包壳 | 304 SS 304 不锈钢 | BWR | Fuel swelling 燃料肿胀 |
| Fuel cladding 燃料包壳 | 304 SS 304 不锈钢 | PWR | Fuel swelling 燃料肿胀 |
| Fuel cladding⁎ 燃料包壳 ⁎ | 20%Cr25%Ni/Nb | AGR | Fuel swelling 燃料肿胀 |
| Fuel cladding ferrules 燃料包壳铁圈 | 20%Cr25%Ni/Nb | SGHWR | Fabrication 制造 |
| Neutron source holders 中子源支架 | 304 SS 304 不锈钢 | BWR | Welding and be swelling 焊接和膨胀 |
| Instrument dry tubes 仪器干管 | 304 SS 304 不锈钢 | BWR | Fabrication 制造 |
| Control rod absorber tubes 控制棒吸收棒 | 304/304L/316L SS 304/304L/316L 不锈钢 | BWR | B4C swelling B 4 C 膨胀 |
| Fuel bundle cap screws 燃料组件螺帽 | 304 SS 304 不锈钢 | BWR | Fabrication 制造 |
| Control rod follower rivets 控制棒跟随铆钉 | 304 SS 304 不锈钢 | BWR | Fabrication 制造 |
| Control blade handle 控制刀柄 | 304 SS 304 不锈钢 | BWR | Low stress 低应力 |
| Control blade sheath 控制叶片套管 | 304 SS 304 不锈钢 | BWR | Low stress 低应力 |
| Control blades 控制叶片 | 304 SS 304 不锈钢 | PWR | Low stress 低应力 |
| Plate type control blade 板式控制叶片 | 304 SS 304 不锈钢 | BWR | Low stress 低应力 |
| Various bolts⁎⁎ 各种螺栓 ⁎⁎ | A-286 | PWR and BWR 压水堆和沸水堆 | Service 服务 |
| Steam separator dryer bolts⁎⁎ 蒸汽分离器干燥螺栓 ⁎⁎ | A286 | BWR | Service 服务 |
| Shroud head bolts⁎⁎ 屏蔽头螺栓 ⁎⁎ | 600 | BWR | Service 服务 |
| Various bolts 各种螺栓 | X-750 | BWR and PWR 沸水堆和压水堆 | Service 服务 |
| Guide tube support pins 导管支撑销 | X-750 | PWR | Service 服务 |
| Jet pump beams 喷气泵束 | X-750 | BWR | Service 服务 |
| Various springs 各种弹簧 | X-750 | BWR and PWR 沸水堆和压水堆 | Service 服务 |
| Various Springs 各种弹簧 | 718 | PWR | Service 服务 |
| Baffle former bolts 挡板前螺栓 | 316 SS cold work 316 不锈钢冷加工 | PWR | Torque, differential swelling 扭矩，差胀 |
| Core shroud 核心屏蔽 | 304/316/347/L SS 304/316/347/L 不锈钢 | BWR | Weld residual stress 焊接残余应力 |
| Top guide 顶部导向 | 304 SS 304 不锈钢 | BWR | Low stress (bending) 低应力（弯曲） |
- ⁎
- Cracking in AGR fuel occurred during storage in spent fuel pool.
AGR 燃料在乏燃料池中储存时发生开裂。 - ⁎⁎
- Cracking of core internals occurred away from high neutron and gamma fluxes.
核心内部构件的开裂发生在高中子和伽马通量之外。
Irradiation causes a significant change in local composition near grain boundaries and other defect sinks [3], [4]. The enrichment of nickel and silicon and the depletion of chromium can affect the susceptibility to IASCC, especially under oxidizing conditions. Irradiation also alters the microstructure and, under LWR conditions, faulty dislocation loops represent the primary irradiation-induced microstructure defect. The loops impede the motion of dislocations, resulting in an increase in the yield strength of stainless steels by as much as a factor of five. Radiation hardening correlates with IASCC propensity, and also induces highly localized deformation in the form of dislocation channels, which could contribute to IASCC [100], [101]. Irradiation also induces creep that can relax macroscopic stresses, and can also enhance local dynamic deformation. Other factors, such as swelling and formation of new phases, may enhance IASCC at high fluence. With the many effects of irradiation that overlap spatially and temporally, more work is needed to identify their roles in the mechanism(s) of IASCC and develop a comprehensive prediction methodology. In particular, the evolution of localized deformation in an irradiation damage microstructure and the emergence of phases at high dose are the areas that require a better understanding to ensure the structural integrity of core components to lives beyond 40 or 60 years. The mechanism of IASCC remains a major outstanding issue in the degradation of core components in LWRs of all types.
辐照会导致晶界和其他缺陷陷阱附近的局部成分发生显著变化[3], [4]。镍和硅的富集以及铬的消耗会影响对辐照-assisted 应力腐蚀开裂（IASCC）的敏感性，尤其是在氧化条件下。辐照还会改变微观结构，在轻水反应堆（LWR）条件下，有缺陷的位错环代表主要的辐照诱导微观结构缺陷。这些环会阻碍位错运动，导致不锈钢的屈服强度增加高达五倍。辐照硬化与 IASCC 倾向相关，并诱导形成位错通道这种高度局部的变形，这可能有助于 IASCC[100], [101]。辐照还会诱导蠕变，可以缓解宏观应力，并可能增强局部动态变形。其他因素，如肿胀和新生相的形成，可能会在高通量下增强 IASCC。由于辐照的许多效应在空间和时间上重叠，需要更多研究来识别它们在 IASCC 机制中的作用，并开发一套综合的预测方法。 特别是在辐照损伤微观结构中局部变形的演化以及高剂量下新相的出现，是确保核心部件在 40 或 60 年以上保持结构完整性的关键领域，需要更深入的理解。IASCC 的机制仍然是各类轻水反应堆核心部件退化中的主要未解决问题。
辐照会导致晶界和其他缺陷陷阱附近的局部成分发生显著变化[3], [4]。镍和硅的富集以及铬的消耗会影响对辐照-assisted 应力腐蚀开裂（IASCC）的敏感性，尤其是在氧化条件下。辐照还会改变微观结构，在轻水反应堆（LWR）条件下，有缺陷的位错环代表主要的辐照诱导微观结构缺陷。这些环会阻碍位错运动，导致不锈钢的屈服强度增加高达五倍。辐照硬化与 IASCC 倾向相关，并诱导形成位错通道这种高度局部的变形，这可能有助于 IASCC[100], [101]。辐照还会诱导蠕变，可以缓解宏观应力，并可能增强局部动态变形。其他因素，如肿胀和新生相的形成，可能会在高通量下增强 IASCC。由于辐照的许多效应在空间和时间上重叠，需要更多研究来识别它们在 IASCC 机制中的作用，并开发一套综合的预测方法。 特别是在辐照损伤微观结构中局部变形的演化以及高剂量下新相的出现，是确保核心部件在 40 或 60 年以上保持结构完整性的关键领域，需要更深入的理解。IASCC 的机制仍然是各类轻水反应堆核心部件退化中的主要未解决问题。
In considering the attributes of an IASCC-resistant austenitic alloy, the following are likely beneficial: high Ni and Cr contents, low Si content, absence of brittle oxide and nitride inclusions, high coincident site lattice fraction of grain boundaries, low connectivity of high-angle grain boundaries and grain boundary coverage by chromium carbides. Finally, ferritic or ferritic–martensitic alloys should be considered due to their inherently greater resistance to radiation effects and IGSCC compared to iron- and nickel-base austenitic alloys [102].
在考虑抗 IASCC 奥氏体合金的属性时，以下特性可能是有益的：高 Ni 和 Cr 含量、低 Si 含量、无脆性氧化物和氮化物夹杂物、高晶界重位点晶格分数、低高角度晶界的连通性以及铬碳化物对晶界的覆盖。最后，应考虑铁素体或铁素体-马氏体合金，因为与铁基和镍基奥氏体合金相比，它们对辐射效应和 IGSCC 具有固有的更高抗性[102]。
在考虑抗 IASCC 奥氏体合金的属性时，以下特性可能是有益的：高 Ni 和 Cr 含量、低 Si 含量、无脆性氧化物和氮化物夹杂物、高晶界重位点晶格分数、低高角度晶界的连通性以及铬碳化物对晶界的覆盖。最后，应考虑铁素体或铁素体-马氏体合金，因为与铁基和镍基奥氏体合金相比，它们对辐射效应和 IGSCC 具有固有的更高抗性[102]。
2.2.2.3. Irradiation-accelerated corrosion
2.2.2.3. 辐照加速腐蚀
Another major issue in LWRs operated to high doses involves the interplay of radiation and corrosion in irradiation-accelerated corrosion. The mechanism of this process is unknown, but it has been shown to have significant impacts on corrosion rates. For example, zirconium irradiated in-reactor in moist carbon dioxide–air mixtures had oxygen weight gains of more than five times that in the unirradiated state [103]. In-reactor corrosion rates of zirconium alloys were found to be 10 times greater than those conducted out-of-reactor, and part of the difference was attributed to greater permeability of the oxide irradiated in-reactor [104]. More recently, Lewis and Hunn [105] found that proton irradiation of a 316 stainless steel foil in room-temperature water for 4 h produced an oxide that was 20 times thicker than the unirradiated control. Further, old data on in-reactor exposure of Zircaloy-2 [106] revealed ten-fold increases in the oxide weight gain and a strong, linear dependence on neutron flux. These data suggest that displacement damage to the solid during corrosion produces a significantly greater effect than that due to radiolysis alone.
在运行于高剂量的轻水反应堆（LWRs）中，另一个主要问题是辐射与腐蚀在辐照加速腐蚀中的相互作用。该过程的机制尚不清楚，但已被证明对腐蚀速率有显著影响。例如，在反应堆内用潮湿的二氧化碳-空气混合物辐照的锆，其氧重量增益比未辐照状态高出五倍以上[103]。研究发现，锆合金的反应堆内腐蚀速率比反应堆外进行的腐蚀速率高 10 倍，其中部分差异归因于反应堆内辐照的氧化物渗透性更大[104]。最近，刘易斯和亨恩[105]发现，将 316 不锈钢箔在室温水中辐照 4 小时后，形成的氧化物比未辐照对照组厚 20 倍。此外，关于 Zircaloy-2 反应堆内辐照的老数据[106]显示，氧化物的重量增益增加了十倍，并且与中子通量呈强线性关系。这些数据表明，腐蚀过程中固体材料的位移损伤产生的效果，比单纯由辐射分解产生的影响要显著得多。
在运行于高剂量的轻水反应堆（LWRs）中，另一个主要问题是辐射与腐蚀在辐照加速腐蚀中的相互作用。该过程的机制尚不清楚，但已被证明对腐蚀速率有显著影响。例如，在反应堆内用潮湿的二氧化碳-空气混合物辐照的锆，其氧重量增益比未辐照状态高出五倍以上[103]。研究发现，锆合金的反应堆内腐蚀速率比反应堆外进行的腐蚀速率高 10 倍，其中部分差异归因于反应堆内辐照的氧化物渗透性更大[104]。最近，刘易斯和亨恩[105]发现，将 316 不锈钢箔在室温水中辐照 4 小时后，形成的氧化物比未辐照对照组厚 20 倍。此外，关于 Zircaloy-2 反应堆内辐照的老数据[106]显示，氧化物的重量增益增加了十倍，并且与中子通量呈强线性关系。这些数据表明，腐蚀过程中固体材料的位移损伤产生的效果，比单纯由辐射分解产生的影响要显著得多。
2.2.3. Reactor pressure vessel integrity issues
2.2.3. 堆芯压力容器完整性问题
The two major degradation issues for RPV steels are embrittlement associated with hardening from radiation-induced solute–defect clusters, and corrosion and stress corrosion cracking phenomena; the corrosion and stress corrosion cracking issues have been briefly summarized in Section 2.2.2.
反应堆压力容器钢的主要两种退化问题是辐射诱导的溶质-缺陷团簇硬化相关的脆化，以及腐蚀和应力腐蚀开裂现象；腐蚀和应力腐蚀开裂问题已在 2.2.2 节中简要概述。
反应堆压力容器钢的主要两种退化问题是辐射诱导的溶质-缺陷团簇硬化相关的脆化，以及腐蚀和应力腐蚀开裂现象；腐蚀和应力腐蚀开裂问题已在 2.2.2 节中简要概述。
Radiation embrittlement is of greatest concern for PWR pressure vessels; the larger vessel diameter in BWRs allows more moderation of fast neutrons in the water between the core and vessel, resulting in 3 to 10 times lower fast neutron fluxes compared to PWR pressure vessels [67]. The fundamental mechanisms associated with radiation hardening and accompanying embrittlement of RPV steels are generally understood, including the pronounced deleterious effects associated with Cu, P and Ni solute additions due to the formation of nanoscale precipitates that produced significant matrix hardening [49], [67], [107]. A series of improved mechanistic models for radiation embrittlement have been developed over the past 25 years that provide a clear description of the complex and synergistic effects associated with solute (Cu, Ni, P), temperature, dose and irradiation flux [49], [108]. Fig. 14 shows an example of the predicted increase in the ductile-to-brittle transition temperature (DBTT) of RPV steel containing high levels of Cu or Ni solute [49]. For the high Cu, medium Ni solute case, a high density of Cu-rich precipitates forms rapidly at low dose and causes significant embrittlement for exposure levels corresponding to a few years of operation of a PWR pressure vessel. A relatively small amount of additional embrittlement occurs after the initial rapid embrittlement phase due to the lack of additional Cu solute. For the high-Ni case, predicted slow nucleation rates of the Ni-rich precipitates result in an extended transient regime with relatively low embrittlement, followed by pronounced embrittlement associated with the copious formation of Mn–Ni–Si precipitates. The full impact of such “late blooming phases” on possible accelerated hardening and embrittlement of reactor vessel steels near the end of their design life is the subject of active current research, and could become a major impediment to additional future extensions in the operating lifetime of PWR power plants.
辐照脆化是压水堆（PWR）压力容器最主要的关注点；沸水堆（BWR）中更大的容器直径使得堆芯与容器之间的水中快中子得到更多 moderation，导致与压水堆压力容器相比，快中子通量降低 3 至 10 倍[67]。与反应堆压力容器（RPV）钢辐照硬化及伴随脆化相关的基本机制通常已被理解，包括由于纳米级沉淀物的形成导致 Cu、P 和 Ni 溶质添加产生显著基体硬化所带来的明显有害效应[49], [67], [107]。在过去 25 年间，已开发了一系列改进的辐照脆化机理模型，这些模型清晰地描述了与溶质（Cu、Ni、P）、温度、剂量和辐照通量相关的复杂和协同效应[49], [108]。图 14 展示了含有高浓度 Cu 或 Ni 溶质的 RPV 钢的延脆转变温度（DBTT）预测增加的一个例子[49]。 对于高 Cu、中 Ni 溶质的情况，在低剂量下会迅速形成高密度的富 Cu 析出物，导致在相当于压水堆压力容器几年运行时间暴露水平下出现显著脆化。由于缺乏额外的 Cu 溶质，在初始快速脆化阶段后会发生相对较小的额外脆化。对于高 Ni 的情况，富 Ni 析出物的预测缓慢形核速率导致一个相对低脆化的扩展瞬态阶段，随后伴随着 Mn-Ni-Si 析出物的大量形成而出现明显的脆化。这种“后期萌发相”对反应堆压力容器钢在其设计寿命末期可能出现的加速硬化与脆化的全面影响是当前活跃的研究课题，并可能成为未来压水堆电站运行寿命进一步延长的重大障碍。
辐照脆化是压水堆（PWR）压力容器最主要的关注点；沸水堆（BWR）中更大的容器直径使得堆芯与容器之间的水中快中子得到更多 moderation，导致与压水堆压力容器相比，快中子通量降低 3 至 10 倍[67]。与反应堆压力容器（RPV）钢辐照硬化及伴随脆化相关的基本机制通常已被理解，包括由于纳米级沉淀物的形成导致 Cu、P 和 Ni 溶质添加产生显著基体硬化所带来的明显有害效应[49], [67], [107]。在过去 25 年间，已开发了一系列改进的辐照脆化机理模型，这些模型清晰地描述了与溶质（Cu、Ni、P）、温度、剂量和辐照通量相关的复杂和协同效应[49], [108]。图 14 展示了含有高浓度 Cu 或 Ni 溶质的 RPV 钢的延脆转变温度（DBTT）预测增加的一个例子[49]。 对于高 Cu、中 Ni 溶质的情况，在低剂量下会迅速形成高密度的富 Cu 析出物，导致在相当于压水堆压力容器几年运行时间暴露水平下出现显著脆化。由于缺乏额外的 Cu 溶质，在初始快速脆化阶段后会发生相对较小的额外脆化。对于高 Ni 的情况，富 Ni 析出物的预测缓慢形核速率导致一个相对低脆化的扩展瞬态阶段，随后伴随着 Mn-Ni-Si 析出物的大量形成而出现明显的脆化。这种“后期萌发相”对反应堆压力容器钢在其设计寿命末期可能出现的加速硬化与脆化的全面影响是当前活跃的研究课题，并可能成为未来压水堆电站运行寿命进一步延长的重大障碍。
Fig. 14. Predicted increase in the radiation-induced DBTT of RPV steel for two solute concentrations [49].
图 14. 两种溶质浓度下 RPV 钢辐射诱导 DBTT 的预测增加[49]
Pressure vessel embrittlement issues that need further research include [49], [67], [108]: (i) the effects of high dose, long operation lifetimes and irradiation flux on hardening and embrittlement; (ii) impact of heat-to-heat variability and the quantitative accuracy of surrogate materials such as surveillance specimens; (iii) effect of Ni concentration on the formation and embrittlement impact of Ni-rich “late blooming phases”; (iv) quantitative validity of the fracture toughness master curve concept (is the curve shape universal, correlation between Charpy impact and fracture toughness tests, consequences of intergranular fracture, etc.); (v) impact of size (constraint) effects and other phenomena on the toughness correlation between pre-cracked Charpy (or smaller) surveillance specimens on possible introduction of biased estimates of reference fracture toughness; (vi) correctly accounting for the large attenuation in neutron fluence and displacement damage dose (a factor of ∼5) that occurs through the wall thickness of reactor vessels; (vii) development of improved-fidelity modeling and microstructural analysis to achieve better understanding of the roles and synergies of the various key experimental parameters; (viii) potential for phosphorus segregation to induce intergranular fracture; (ix) thermal annealing and reirradiation research to investigate feasibility and quantify optimized conditions for periodic vessel annealing to mitigate radiation embrittlement; and (x) investigation of long-term (>50 years) thermal aging effects on the microstructure and properties of low alloy steels. Finally, research on neutron radiation embrittlement of weldments (including the impact of various post weld heat treatments) should be continued.
需要进一步研究的压力容器脆化问题包括[49]、[67]、[108]：(i)高剂量、长运行寿命和辐照通量对硬化与脆化的影响；(ii)热致热变化的影响以及替代材料（如监测试样）的定量准确性；(iii)Ni 浓度对富 Ni“后期形成相”形成与脆化影响的作用；(iv)断裂韧性主曲线概念的定量有效性（曲线形状是否普遍、夏比冲击与断裂韧性测试的相关性、晶间断裂的后果等）；(v)尺寸（约束）效应及其他现象对预裂纹夏比（或更小）监测试样韧性相关性的影响，可能导致参考断裂韧性的偏倚估计；(vi)正确考虑反应堆容器壁厚方向中中子注量率和位移损伤剂量的大幅衰减（约 5 倍）；(vii)开发更高保真度的建模和微观结构分析，以更好地理解各种关键实验参数的作用和协同效应；(viii)磷偏析导致晶间断裂的可能性；(ix)热退火和再辐照研究，以探究周期性容器退火的可行性与优化条件，以缓解辐照脆化；(x)研究长期（>50 年）热老化对低合金钢微观结构和性能的影响。 最后，应继续研究焊缝的中子辐射脆化（包括各种焊后热处理的影响）。
需要进一步研究的压力容器脆化问题包括[49]、[67]、[108]：(i)高剂量、长运行寿命和辐照通量对硬化与脆化的影响；(ii)热致热变化的影响以及替代材料（如监测试样）的定量准确性；(iii)Ni 浓度对富 Ni“后期形成相”形成与脆化影响的作用；(iv)断裂韧性主曲线概念的定量有效性（曲线形状是否普遍、夏比冲击与断裂韧性测试的相关性、晶间断裂的后果等）；(v)尺寸（约束）效应及其他现象对预裂纹夏比（或更小）监测试样韧性相关性的影响，可能导致参考断裂韧性的偏倚估计；(vi)正确考虑反应堆容器壁厚方向中中子注量率和位移损伤剂量的大幅衰减（约 5 倍）；(vii)开发更高保真度的建模和微观结构分析，以更好地理解各种关键实验参数的作用和协同效应；(viii)磷偏析导致晶间断裂的可能性；(ix)热退火和再辐照研究，以探究周期性容器退火的可行性与优化条件，以缓解辐照脆化；(x)研究长期（>50 年）热老化对低合金钢微观结构和性能的影响。 最后，应继续研究焊缝的中子辐射脆化（包括各种焊后热处理的影响）。
2.3. Materials challenges during off-normal events
2.3. 非正常事件期间的材料挑战
There are two major transient operational scenarios that have been considered for design basis safety analyses used to define operational safety limits for commercial reactors: reactivity-initiated accidents (RIAs) and loss-of-cooling accidents (LOCAs). RIA events are an unwanted increase in fission rate (reactor power), such as might occur from an unintended prompt ejection of a control rod [109]. The RIA scenario would produce a rapid increase in fuel power and temperature, which could lead to potential failure of fuel cladding and release of radioactivity into the primary coolant. Due to degradation of the cladding mechanical properties with increasing exposure, safety authorities have established design limits for the maximum allowable energy deposition in the fuel, which decrease with increasing burn-up. The RIA event would be relatively short in duration since the negative coefficient of reactivity in LWRs would quickly lead to a reduction of the excess reactivity and primary coolant flow would continue throughout the event. The LOCA scenarios are based on the knowledge that fission reactors continue to generate substantial amounts of decay heat for days and weeks after they are shut down, due to a variety of radioactivity decay processes in the fuel and core materials; for example, the residual heat relative to the full-power reactor value is about 6% at a time of 1 s after a reactor scram and ∼0.5% 1 day after shutdown. Considering that commercial reactors typically create over 3000 MW of thermal power at full power, these residual heat values represent significant heat removal challenges for a LOCA scenario. Without adequate heat removal, the increase in core temperature can lead to cladding rupture and release of radioactivity into the primary coolant. The LOCA event could last for hours or days at elevated temperature, depending on the severity of the loss of cooling capability.
在用于定义商业反应堆运行安全限值的基准安全分析中，已考虑了两种主要的瞬态运行场景：反应性引发事故（RIAs）和失冷事故（LOCA）。RIA 事件是指裂变率（反应堆功率）的意外增加，例如控制棒意外弹出[109]可能引发的情况。RIA 场景会导致燃料功率和温度的快速增加，这可能导致燃料包壳的潜在失效，并使放射性物质释放到一回路冷却剂中。由于包壳的机械性能随辐照时间的增加而退化，安全当局已为燃料中允许的最大能量沉积建立了设计限值，该限值随燃耗的增加而降低。由于轻水反应堆的反应性负系数会迅速导致过剩反应性的减少，并且在整个事件期间一回路冷却剂流量会持续，因此 RIA 事件持续时间相对较短。 小破口事故场景基于以下知识：裂变反应堆在关闭后数天和数周内仍会因燃料和核心材料中的各种放射性衰变过程继续产生大量衰变热；例如，反应堆紧急停堆后 1 秒时的残余热量相对于满功率反应堆值约为 6%，停堆 1 天后约为 0.5%。考虑到商业反应堆在满功率时通常会产生超过 3000 兆瓦的热功率，这些残余热量值对 LOCA 场景构成了显著的热量移除挑战。如果没有足够的热量移除，核心温度的升高会导致包壳破裂，放射性物质释放到一回路冷却剂中。LOCA 事件可能在高温下持续数小时或数天，具体取决于冷却能力丧失的严重程度。
在用于定义商业反应堆运行安全限值的基准安全分析中，已考虑了两种主要的瞬态运行场景：反应性引发事故（RIAs）和失冷事故（LOCA）。RIA 事件是指裂变率（反应堆功率）的意外增加，例如控制棒意外弹出[109]可能引发的情况。RIA 场景会导致燃料功率和温度的快速增加，这可能导致燃料包壳的潜在失效，并使放射性物质释放到一回路冷却剂中。由于包壳的机械性能随辐照时间的增加而退化，安全当局已为燃料中允许的最大能量沉积建立了设计限值，该限值随燃耗的增加而降低。由于轻水反应堆的反应性负系数会迅速导致过剩反应性的减少，并且在整个事件期间一回路冷却剂流量会持续，因此 RIA 事件持续时间相对较短。 小破口事故场景基于以下知识：裂变反应堆在关闭后数天和数周内仍会因燃料和核心材料中的各种放射性衰变过程继续产生大量衰变热；例如，反应堆紧急停堆后 1 秒时的残余热量相对于满功率反应堆值约为 6%，停堆 1 天后约为 0.5%。考虑到商业反应堆在满功率时通常会产生超过 3000 兆瓦的热功率，这些残余热量值对 LOCA 场景构成了显著的热量移除挑战。如果没有足够的热量移除，核心温度的升高会导致包壳破裂，放射性物质释放到一回路冷却剂中。LOCA 事件可能在高温下持续数小时或数天，具体取决于冷却能力丧失的严重程度。
Whereas zirconium alloys have evolved to achieve very impressive fuel performance under normal operating conditions in water-cooled nuclear reactors (Section 2.2.1), the behavior of UO2 monolithic fuel with Zr alloy cladding under accident scenarios is far from optimized. Hydrogen uptake and radiation-induced hardening can lead to reduced ductility in Zr alloy cladding that can impair the resistance to cladding failure associated with pellet–cladding mechanical interactions during an RIA event (expansion of the fuel pellet due to rapid heating) [110]. If pronounced clad breaching occurs, the possibility exists for significant fragmentation and dispersal of the ceramic UO2 fuel pellets. For LOCA scenarios, the rapid increase in Zr alloy oxidation rate with increasing temperature will degrade the mechanical properties of the cladding and could lead to cladding breach and/or fracture, which in turn could lead to reduced flow of primary coolant within the fuel assemblies in the core due to coolant channel blockage by cladding fragments. The high heat of oxidation for Zr can also make a large contribution to the core heating; considering that a typical 1000 MWe LWR core contains ∼30,000 kg of cladding, ∼55 MW h of heat would be released if it were completely oxidized (note: BWR cores also contain nearly a comparable amount of Zr in the channel boxes). This could set up an autocatalytic reaction where the heat from Zr oxidation drives a temperature increase in the core, leading to more rapid oxidation and heat generation. Finally, the oxidation of Zr by steam leads to the production of potentially explosive hydrogen gas (∼1200 kg if all of the cladding was oxidized by steam). The shortcomings of the Zr alloy/UO2 fuel system to severe accident conditions such as occurred at the Fukushima Dai-ichi nuclear power plant in Japan following the 2011 earthquake and tsunami has led to increased interest in improving the safety of nuclear reactors to rare but credible accident scenarios [111].
尽管锆合金在轻水核反应堆的常规运行条件下已发展出非常优异的燃料性能（第 2.2.1 节），但在事故场景下，带有锆合金包壳的 UO 2 整体燃料的行为远未得到优化。氢的吸收和辐射诱导硬化会导致锆合金包壳的延展性降低，从而损害 RIA 事件（燃料颗粒因快速加热而膨胀）期间颗粒-包壳机械相互作用相关的包壳失效抗力[110]。如果发生明显的包壳破损，陶瓷 UO 2 燃料颗粒存在显著破碎和散播的可能性。对于失水事故场景，锆合金氧化速率随温度升高而迅速增加，这将降低包壳的力学性能，并可能导致包壳破损和/或断裂，进而由于包壳碎片堵塞冷却剂通道，导致反应堆芯内燃料组件的冷却剂流量减少。 锆的氧化热高也会对堆芯加热产生很大影响；考虑到一个典型的 1000 兆瓦压水堆堆芯含有约 30,000 公斤的包壳，如果完全氧化，将释放约 55 兆瓦时的热量（注：沸水堆堆芯的通道箱中也含有几乎相当的锆）。这可能会引发自催化反应，其中锆氧化产生的热量导致堆芯温度升高，进而导致更快的氧化和热量产生。最后，蒸汽氧化锆会产生潜在的爆炸性氢气（如果所有包壳都被蒸汽氧化，将产生约 1200 公斤）。锆合金/UO 燃料系统在严重事故条件下的不足，例如 2011 年日本地震和海啸后在福岛第一核电站发生的事故，已导致人们对提高核反应堆在罕见但可信的事故场景中的安全性产生了浓厚兴趣[111]。
尽管锆合金在轻水核反应堆的常规运行条件下已发展出非常优异的燃料性能（第 2.2.1 节），但在事故场景下，带有锆合金包壳的 UO 2 整体燃料的行为远未得到优化。氢的吸收和辐射诱导硬化会导致锆合金包壳的延展性降低，从而损害 RIA 事件（燃料颗粒因快速加热而膨胀）期间颗粒-包壳机械相互作用相关的包壳失效抗力[110]。如果发生明显的包壳破损，陶瓷 UO 2 燃料颗粒存在显著破碎和散播的可能性。对于失水事故场景，锆合金氧化速率随温度升高而迅速增加，这将降低包壳的力学性能，并可能导致包壳破损和/或断裂，进而由于包壳碎片堵塞冷却剂通道，导致反应堆芯内燃料组件的冷却剂流量减少。 锆的氧化热高也会对堆芯加热产生很大影响；考虑到一个典型的 1000 兆瓦压水堆堆芯含有约 30,000 公斤的包壳，如果完全氧化，将释放约 55 兆瓦时的热量（注：沸水堆堆芯的通道箱中也含有几乎相当的锆）。这可能会引发自催化反应，其中锆氧化产生的热量导致堆芯温度升高，进而导致更快的氧化和热量产生。最后，蒸汽氧化锆会产生潜在的爆炸性氢气（如果所有包壳都被蒸汽氧化，将产生约 1200 公斤）。锆合金/UO 燃料系统在严重事故条件下的不足，例如 2011 年日本地震和海啸后在福岛第一核电站发生的事故，已导致人们对提高核反应堆在罕见但可信的事故场景中的安全性产生了浓厚兴趣[111]。
Considering that the key functions of the fuel system in an accident scenario are to maintain core cooling capability and to minimize or prevent dispersal of fuel and fission products, there are three major potential approaches to design LWR fuel systems with improved accident tolerance. (i) Utilize cladding options with reduced reaction kinetics with high-temperature steam; reduction of the cladding high-temperature oxidation rate by one or more orders of magnitude compared to alloys such as Zircaloy would nearly eliminate the contribution of heat input from oxidation and would proportionally reduce the generation of hydrogen. Possible approaches in this category include utilization of oxidation-resistant coatings or development of new alloys with significantly improved oxidation resistance (either Zr-based or new alloys). (ii) Utilize fuel cladding with improved high-temperature mechanical properties and resistance to hydrogen embrittlement (high tensile and short-term creep strength, good thermal shock resistance, high melting temperature). (iii) Utilize new fuel forms with lower operating temperature, higher margins to fuel melting, enhanced retention of fission products (e.g. fuel microencapsulation or gettering techniques), and/or reduced fuel dispersal probability compared to UO2. Since fuel systems are inherently a replaceable commodity, it would be possible to implement new fuels with incremental improvements in accident tolerance (using one or more of the approaches outlined above) as long as the geometry and performance under normal operations do not have a negative impact on the reactor operations.
考虑到在事故场景中，燃料系统的关键功能是维持堆芯冷却能力，并尽量减少或防止燃料和裂变产物的扩散，因此设计具有改进事故耐受性的轻水堆燃料系统主要有三种潜在方法。(i) 利用与高温蒸汽反应动力学降低的包壳选项；与锆合金等合金相比，将包壳高温氧化速率降低一个或多个数量级，将几乎消除氧化热输入的贡献，并按比例减少氢气的产生。此类方法包括利用抗氧化涂层或开发具有显著改善抗氧化性能的新合金（无论是锆基合金还是新合金）。(ii) 利用具有改进高温力学性能和抗氢脆性的燃料包壳（高强度和短期蠕变强度、良好的热冲击抗性、高熔点）。 (iii) 利用较低温运行的新型燃料形式，具有更高的燃料熔化裕度、增强的裂变产物（例如燃料微封装或吸气技术）的保留能力，以及与 UO₂相比更低的燃料分散概率。由于燃料系统本质上是一种可替换的商品，只要几何形状和正常运行下的性能不会对反应堆运行产生负面影响，就可以通过（使用上述一种或多种方法）以增量改进的方式实施新型燃料，提高事故耐受性。
考虑到在事故场景中，燃料系统的关键功能是维持堆芯冷却能力，并尽量减少或防止燃料和裂变产物的扩散，因此设计具有改进事故耐受性的轻水堆燃料系统主要有三种潜在方法。(i) 利用与高温蒸汽反应动力学降低的包壳选项；与锆合金等合金相比，将包壳高温氧化速率降低一个或多个数量级，将几乎消除氧化热输入的贡献，并按比例减少氢气的产生。此类方法包括利用抗氧化涂层或开发具有显著改善抗氧化性能的新合金（无论是锆基合金还是新合金）。(ii) 利用具有改进高温力学性能和抗氢脆性的燃料包壳（高强度和短期蠕变强度、良好的热冲击抗性、高熔点）。 (iii) 利用较低温运行的新型燃料形式，具有更高的燃料熔化裕度、增强的裂变产物（例如燃料微封装或吸气技术）的保留能力，以及与 UO₂相比更低的燃料分散概率。由于燃料系统本质上是一种可替换的商品，只要几何形状和正常运行下的性能不会对反应堆运行产生负面影响，就可以通过（使用上述一种或多种方法）以增量改进的方式实施新型燃料，提高事故耐受性。
In addition to the potential changes to the fuel system, there are several other core materials systems that would be considered for improvements if robust accident tolerant fuels were successfully developed. For example, the AgInCd control rods currently used in many PWRs have a relatively low melting temperature (∼900 °C). Similarly, the control blades in many BWRs, which utilize B4C pellets inside stainless steel tubes, begin to have a eutectic Fe–B reaction at ∼1170 °C. Higher-temperature material options for neutron control should be pursued.
除了燃料系统的潜在变化外，如果成功开发出具有高事故耐受性的燃料，还有几个其他核心材料系统将考虑进行改进。例如，目前许多压水堆使用的 AgInCd 控制棒具有相对较低的熔点（约 900 °C）。类似地，许多沸水堆中使用的控制叶片，其内部嵌有不锈钢管内的 B₄C 颗粒，在约 1170 °C 时开始发生共晶 Fe-B 反应。应追求用于中子控制的更高温度材料选项。
除了燃料系统的潜在变化外，如果成功开发出具有高事故耐受性的燃料，还有几个其他核心材料系统将考虑进行改进。例如，目前许多压水堆使用的 AgInCd 控制棒具有相对较低的熔点（约 900 °C）。类似地，许多沸水堆中使用的控制叶片，其内部嵌有不锈钢管内的 B₄C 颗粒，在约 1170 °C 时开始发生共晶 Fe-B 反应。应追求用于中子控制的更高温度材料选项。
3. Materials challenges in future fission reactor concepts
3. 未来裂变反应堆概念中的材料挑战
Construction is currently in progress worldwide on several so-called Generation III and Generation III+ LWR power plants that are designed for improved efficiency, passive safety and economics. To a large extent, these reactors represent an evolutionary design change utilizing materials systems that are similar to current (Generation II) LWRs, and therefore the materials challenges facing the new reactors will be comparable to those faced in existing reactors. Another class of light-water-cooled reactors under consideration would utilize in-factory construction techniques and new designs with high emphasis on passive safety to construct small (50–300 MWe) nuclear power plants; some of the materials challenges with these small modular reactors (SMRs) are discussed in the following section. Finally, a brief discussion on the materials challenges for Generation IV reactor concepts will be provided.
目前，全球正在建设几座所谓的第三代和第三代+轻水反应堆核电站，这些核电站的设计旨在提高效率、被动安全性和经济性。在很大程度上，这些反应堆代表了一种渐进式的设计变化，利用的材料系统与当前的（第二代）轻水反应堆相似，因此新反应堆面临的材料挑战将与现有反应堆面临的挑战相当。另一类考虑中的轻水冷却反应堆将采用工厂内建造技术和新的设计，高度强调被动安全性，以建造小型（50–300 MW）核电站；这些小型模块化反应堆（SMRs）的部分材料挑战将在下一节中讨论。最后，将对第四代反应堆概念的材料挑战进行简要讨论。
目前，全球正在建设几座所谓的第三代和第三代+轻水反应堆核电站，这些核电站的设计旨在提高效率、被动安全性和经济性。在很大程度上，这些反应堆代表了一种渐进式的设计变化，利用的材料系统与当前的（第二代）轻水反应堆相似，因此新反应堆面临的材料挑战将与现有反应堆面临的挑战相当。另一类考虑中的轻水冷却反应堆将采用工厂内建造技术和新的设计，高度强调被动安全性，以建造小型（50–300 MW）核电站；这些小型模块化反应堆（SMRs）的部分材料挑战将在下一节中讨论。最后，将对第四代反应堆概念的材料挑战进行简要讨论。
3.1. Materials challenges for proposed light-water small modular reactors
3.1. 提议中的轻水小型模块化反应堆的材料挑战
Due to the high capital cost for constructing a large (∼1000 MWe) nuclear power station, a variety of designs for small modular reactors (<300 MWe) have been proposed [112]. In order to compete against the traditional economies of scale, which favor large reactor size (per MWe), the proposed small reactors would utilize modular construction techniques in factories followed by shipping of major components by rail or barge for final construction at the reactor site. The conceptual designs include both light water and other coolants [112], [113]. The materials issues facing the advanced reactor concepts (coolants other than light water) are qualitatively similar to those for conceptual Generation IV systems discussed in Section 3.2. Many of the specific SMR materials issues are uncertain due to the preliminary nature of their engineering designs at this time. For example, it is uncertain whether there will be any significant modification in the water chemistry for the SMRs compared to what has been utilized in existing LWRs.
由于建设大型（约 1000 MW e ）核电站需要高昂的资本成本，因此已经提出了多种小型模块化反应堆（<300 MW e ）的设计方案[112]。为了与传统的规模经济（有利于大型反应堆，每兆瓦 e ）竞争，提出的小型反应堆将采用工厂内的模块化建造技术，然后通过铁路或驳船运输主要部件，在反应堆现场进行最终建设。概念设计包括轻水和其他冷却剂[112]、[113]。先进反应堆概念（非轻水冷却剂）面临的材料问题，在性质上与第 3.2 节中讨论的概念第四代系统相似。由于这些工程设计的初步性质，许多具体的 SMR 材料问题目前尚不确定。例如，目前尚不确定 SMR 的水化学是否与现有 LWRs 中使用的相比会有任何重大变化。
由于建设大型（约 1000 MW e ）核电站需要高昂的资本成本，因此已经提出了多种小型模块化反应堆（<300 MW e ）的设计方案[112]。为了与传统的规模经济（有利于大型反应堆，每兆瓦 e ）竞争，提出的小型反应堆将采用工厂内的模块化建造技术，然后通过铁路或驳船运输主要部件，在反应堆现场进行最终建设。概念设计包括轻水和其他冷却剂[112]、[113]。先进反应堆概念（非轻水冷却剂）面临的材料问题，在性质上与第 3.2 节中讨论的概念第四代系统相似。由于这些工程设计的初步性质，许多具体的 SMR 材料问题目前尚不确定。例如，目前尚不确定 SMR 的水化学是否与现有 LWRs 中使用的相比会有任何重大变化。
For the light-water-cooled designs that are being proposed for construction in the near future (within 10 years), most of the materials issues are expected to be comparable to those faced by Generation II and Generation III reactors. One key issue will be to determine if the manufacturing processes used in the factory assembly provide any advantage (or disadvantage) in material performance compared to standard fabrication techniques. For example, advanced manufacturing techniques such as additive manufacturing could potentially reduce the fabrication costs of specialized pump components and other devices, but it would be important to compare the microstructure and performance of the materials to conventional wrought material under prototypic operating environments. One of the key design changes in many of the light-water small modular reactors is the use of integral containment (i.e. the steam generator for a PWR is located within the RPV). In some concepts, the primary water flows across the outside of the tubes, resulting in a reversal in primary and secondary sides of conventional steam generators used in all operating PWRs. Integral containment has a significant safety advantage in that the possibility of a large-pipe LOCA is eliminated. However, considering that nuclear reactors still occasionally experience stress corrosion cracking issues with steam generator tubing, and that the evolution of water chemistry control is strongly coupled to the steam generator geometry, the inversion of primary and secondary sides of a tubed steam generator and the limited accessibility of the steam generator for routine inspection may be a source of problems unless further improvements in steam generator reliability are achieved.
对于计划在未来 10 年内建造的轻水冷却设计，大多数材料问题预计与第二代和第三代反应堆面临的问题相当。一个关键问题是确定工厂组装中使用的制造工艺是否在材料性能方面相对于标准制造技术有任何优势（或劣势）。例如，增材制造等先进制造技术有可能降低专用泵部件和其他设备的制造成本，但重要的是要在原型运行环境下比较材料的微观结构和性能与常规锻造材料的差异。许多轻水小型模块化反应堆的一个关键设计变化是采用整体包容式设计（即压水堆的蒸汽发生器位于反应堆压力容器内）。在某些概念中，一回路水流经管子外部，导致所有运行中的压水堆所使用的传统蒸汽发生器的一回路和二回路发生逆转。 整体包容系统具有显著的安全优势，因为它消除了大管道失水事故的可能性。然而，考虑到核反应堆仍然偶尔会因蒸汽发生器管路出现应力腐蚀开裂问题，并且水化学控制的发展与蒸汽发生器几何形状密切相关，除非蒸汽发生器可靠性得到进一步改进，否则管式蒸汽发生器的一、二次侧倒置以及蒸汽发生器日常检查的可及性有限，可能会成为问题的根源。
对于计划在未来 10 年内建造的轻水冷却设计，大多数材料问题预计与第二代和第三代反应堆面临的问题相当。一个关键问题是确定工厂组装中使用的制造工艺是否在材料性能方面相对于标准制造技术有任何优势（或劣势）。例如，增材制造等先进制造技术有可能降低专用泵部件和其他设备的制造成本，但重要的是要在原型运行环境下比较材料的微观结构和性能与常规锻造材料的差异。许多轻水小型模块化反应堆的一个关键设计变化是采用整体包容式设计（即压水堆的蒸汽发生器位于反应堆压力容器内）。在某些概念中，一回路水流经管子外部，导致所有运行中的压水堆所使用的传统蒸汽发生器的一回路和二回路发生逆转。 整体包容系统具有显著的安全优势，因为它消除了大管道失水事故的可能性。然而，考虑到核反应堆仍然偶尔会因蒸汽发生器管路出现应力腐蚀开裂问题，并且水化学控制的发展与蒸汽发生器几何形状密切相关，除非蒸汽发生器可靠性得到进一步改进，否则管式蒸汽发生器的一、二次侧倒置以及蒸汽发生器日常检查的可及性有限，可能会成为问题的根源。
3.2. Proposed next-generation (Generation IV) fission reactor concepts
3.2. 拟议的下一代（第四代）裂变反应堆概念
3.2.1. Brief overview of the six Generation IV concepts
3.2.1. 六种第四代概念简介
Over the past 10 years, the United States Department of Energy and the Generation IV International Forum have explored six particularly appealing advanced reactor concepts as potential next-generation (Generation IV) nuclear power systems [114], [115]. These concepts were selected from hundreds of ideas submitted to the US DOE by scientists and engineers worldwide, during a broad canvassing operation as part of the first phase of the collaborative Generation IV program in 2002 [114]. The objective was to identify concepts that had one or more of the following attributes: increased efficiency, generation of process heat to drive chemical processes such as the production of hydrogen, increased safety and reduction in waste generation. The concepts finally selected were the supercritical-water-cooled reactor (SCWR), the sodium fast reactor (SFR), the lead fast reactor (LFR), the very-high-temperature reactor (VHTR), the gas fast reactor (GFR) and the molten salt reactor (MSR). Table 3 summarizes the basic characteristics of each of these reactor types and the materials proposed for the various major components [7]. Note that all designs call for higher operating temperatures and radiation doses, placing a higher burden on the integrity of materials.
在过去 10 年里，美国能源部与第四代国际论坛共同探索了六种特别有吸引力的先进反应堆概念，作为潜在的下一代（第四代）核动力系统[114]，[115]。这些概念是从 2002 年第四代合作计划第一阶段中，在全球科学家和工程师向美国能源部提交的数百个想法中选出的[114]。目标是识别具有以下一项或多项特性的概念：提高效率、产生工艺热以驱动化学过程（如氢气生产）、提高安全性和减少废物产生。最终选定的概念包括超临界水冷反应堆（SCWR）、钠冷快堆（SFR）、铅冷快堆（LFR）、超高温反应堆（VHTR）、气冷快堆（GFR）和熔盐反应堆（MSR）。表 3 总结了这些反应堆类型的基本特性以及为各种主要部件提出的材料[7]。 请注意，所有设计都要求更高的运行温度和辐射剂量，这对材料的完整性提出了更高的要求。
在过去 10 年里，美国能源部与第四代国际论坛共同探索了六种特别有吸引力的先进反应堆概念，作为潜在的下一代（第四代）核动力系统[114]，[115]。这些概念是从 2002 年第四代合作计划第一阶段中，在全球科学家和工程师向美国能源部提交的数百个想法中选出的[114]。目标是识别具有以下一项或多项特性的概念：提高效率、产生工艺热以驱动化学过程（如氢气生产）、提高安全性和减少废物产生。最终选定的概念包括超临界水冷反应堆（SCWR）、钠冷快堆（SFR）、铅冷快堆（LFR）、超高温反应堆（VHTR）、气冷快堆（GFR）和熔盐反应堆（MSR）。表 3 总结了这些反应堆类型的基本特性以及为各种主要部件提出的材料[7]。 请注意，所有设计都要求更高的运行温度和辐射剂量，这对材料的完整性提出了更高的要求。
To allow operation at much higher temperatures, advanced Generation IV reactor concepts utilize different coolants, including water in the supercritical state, liquid metals such as sodium and lead–bismuth, molten salts and high-pressure helium gas. The materials challenges for the Generation IV reactor concepts come about because of the very high fuel temperatures, the intense radiation flux and coolant compatibility issues. Thus, the fuel, the cladding, the structural materials, the reactor vessel and the interaction of these materials with the coolants present the greatest challenges to new, more robust nuclear reactor concepts for the twenty-first century. Structural materials used in the cores of advanced reactors will face unprecedented combinations of temperature, radiation dose and stress. As shown in Fig. 15, a common feature of all advanced designs is a high operating temperature compared to current LWRs. Another unique feature is the simultaneous presence of intense knock-on displacement damage by the fission neutrons. Almost all of the Generation IV concepts call for radiation damage levels that exceed those of LWR experience. One additional proposed concept, the traveling wave reactor (TWR), would require fuel cladding integrity to damage levels approaching 600 dpa. The need for more robust materials extends to the pressure vessel, the primary safety structure for most reactor designs.
为了允许在更高的温度下运行，先进的第四代反应堆概念采用了不同的冷却剂，包括超临界状态的水、钠和铅铋等液态金属、熔融盐以及高压氦气。第四代反应堆概念的材料挑战源于极高的燃料温度、强烈的辐射通量以及冷却剂兼容性问题。因此，燃料、包壳、结构材料、反应堆容器以及这些材料与冷却剂之间的相互作用，为 21 世纪更坚固的核反应堆概念带来了最大的挑战。先进反应堆核心中使用的结构材料将面临前所未有的温度、辐射剂量和应力组合。如图 15 所示，所有先进设计的共同特点是相比当前的轻水反应堆，具有更高的运行温度。另一个独特特征是裂变中子同时存在强烈的碰撞位移损伤。几乎所有的第四代反应堆概念都要求辐射损伤水平超过轻水反应堆的运行经验。 另一个提议的概念是行波反应堆（TWR），它需要燃料包壳的完整性达到接近 600 dpa 的损伤水平。对更坚固材料的需求扩展到了压力容器，它是大多数反应堆设计中的主要安全结构。
为了允许在更高的温度下运行，先进的第四代反应堆概念采用了不同的冷却剂，包括超临界状态的水、钠和铅铋等液态金属、熔融盐以及高压氦气。第四代反应堆概念的材料挑战源于极高的燃料温度、强烈的辐射通量以及冷却剂兼容性问题。因此，燃料、包壳、结构材料、反应堆容器以及这些材料与冷却剂之间的相互作用，为 21 世纪更坚固的核反应堆概念带来了最大的挑战。先进反应堆核心中使用的结构材料将面临前所未有的温度、辐射剂量和应力组合。如图 15 所示，所有先进设计的共同特点是相比当前的轻水反应堆，具有更高的运行温度。另一个独特特征是裂变中子同时存在强烈的碰撞位移损伤。几乎所有的第四代反应堆概念都要求辐射损伤水平超过轻水反应堆的运行经验。 另一个提议的概念是行波反应堆（TWR），它需要燃料包壳的完整性达到接近 600 dpa 的损伤水平。对更坚固材料的需求扩展到了压力容器，它是大多数反应堆设计中的主要安全结构。
Fig. 15. Temperature and dose requirements for in-core structural materials for the operation of the six proposed Generation IV advanced reactor concepts, the traveling wave reactor and fusion reactor concepts. The dimensions of the colored rectangles represent the ranges of temperature and displacement damage for each reactor concept.
图 15. 六种提议的第四代先进反应堆概念、行波反应堆和聚变反应堆概念中，用于核心结构材料运行的温度和剂量要求。彩色矩形的大小表示每个反应堆概念的温度和位移损伤范围。
3.2.2. Materials-limiting phenomena for Generation IV concepts
3.2.2. 第四代核能概念中的材料限制现象
The high temperatures and damage levels of all of the designs described in Table 3 will accelerate the corrosion and oxidation kinetics and open new pathways for materials degradation. The six Generation IV concepts and the TWR can be grouped into three general categories according to their positions in Fig. 15, as well as the nature of the coolant. The SFR, LFR, MSR and the TWR will all operate at elevated temperatures and very high damage levels, and will utilize either liquid metal or molten salt as the coolant. The GFR and VHTR will experience lower damage levels but even higher temperatures, and will use the relatively innocuous He as the working fluid. The SCWR is a unique third category in that it is the only water-cooled reactor among the Generation IV designs; it will experience lower damage levels relative to other concepts (comparable damage to current LWRs), but conversely will operate at high temperature and very high pressure, where corrosion and stress corrosion cracking issues may become paramount.
表 3 中所有设计的较高温度和损伤水平将加速腐蚀和氧化动力学，并为材料退化开辟新的途径。根据图 15 中的位置以及冷却剂的性质，六种第四代核能概念和 TWR 可分为三大类。SFR、LFR、MSR 和 TWR 都将运行在较高温度和非常高的损伤水平下，并将使用液态金属或熔盐作为冷却剂。GFR 和 VHTR 将经历较低的损伤水平，但温度更高，并将使用相对无害的氦气作为工作流体。SCWR 是独特的第三类，因为它是第四代设计中唯一的用水冷却的反应堆；与其他概念相比，它将经历较低的损伤水平（与当前 LWRs 相当的损伤），但反之将运行在高温和高压下，此时腐蚀和应力腐蚀开裂问题可能变得至关重要。
表 3 中所有设计的较高温度和损伤水平将加速腐蚀和氧化动力学，并为材料退化开辟新的途径。根据图 15 中的位置以及冷却剂的性质，六种第四代核能概念和 TWR 可分为三大类。SFR、LFR、MSR 和 TWR 都将运行在较高温度和非常高的损伤水平下，并将使用液态金属或熔盐作为冷却剂。GFR 和 VHTR 将经历较低的损伤水平，但温度更高，并将使用相对无害的氦气作为工作流体。SCWR 是独特的第三类，因为它是第四代设计中唯一的用水冷却的反应堆；与其他概念相比，它将经历较低的损伤水平（与当前 LWRs 相当的损伤），但反之将运行在高温和高压下，此时腐蚀和应力腐蚀开裂问题可能变得至关重要。
3.2.2.1. High-temperature, high-dose fission concepts
3.2.2.1. 高温、高剂量裂变概念
The challenges associated with the high-temperature, high-dose operating environment of the first group (SFR, LFR, MSR, TWR) will place increased emphasis on strength, creep and creep-fatigue behavior in addition to fracture toughness at low temperature. While chemical interaction between coolant and structural materials will present some degradation challenges, the major concern is the very high radiation damage levels expected in core components. At such high damage levels, the major degradation modes are likely to be driven by void swelling and phase stability. Void swelling occurs at homologous temperatures of 0.35–0.55TM, which for steels (325–650 °C) overlaps the temperature range of the reactor core for these four high-temperature, high-dose concepts. Fig. 16 compares the void swelling behavior (obtained from immersion density measurements) for Type 304L [116], 316 [12] and a Ti-modified (D9) [12] austenitic stainless steel and 9–12% Cr-tempered ferritic/martensitic steels [12], [117], [118], [119] following irradiation at ∼400–550 °C to high doses in a fast fission reactor spectrum. In all cases, the void swelling behavior consists of an initial low-swelling transient regime followed by a high swelling rate regime (approaching 1% dpa−1 for the austenitic steels [12]). Although the maximum allowable volumetric swelling for structural applications is design dependent, void swelling levels >5% are generally unacceptable based on typical engineering design considerations. Severe embrittlement has also been observed in irradiated austenitic steels when the volumetric swelling is >10% [120]. Many years of research and development of austenitic alloys have managed to extend the low-swelling transient regime by utilizing swelling resistant microstructures such as the fine TiC precipitates in Ti-modified austenitic steels, but the delay of steady-state void growth is not sufficient to avoid significant void swelling during operation to the high doses contemplated for several Generation IV reactor concepts (cf. Fig. 9, Fig. 10). As such, more radiation-resistant alloys, such as 2.5–12% Cr bainitic–ferritic–martensitic steels are being considered for high-dose core internal and RPV applications. Exacerbating the problem is phase instability at high doses due to radiation-induced or -enhanced solute segregation [121], [122] and ballistic dissolution of precipitates [122] by energetic displacement cascades. Irradiation can nucleate or dissolve phases, changing the solute composition of the matrix and enhancing void growth [123]. Further, dissolution of particles added to increase the strength of the alloy results in softening and compromises high-temperature strength and creep. For example, γ′ matrix precipitates that provide strength to nickel-base alloys used in high-temperature applications are unstable under irradiation [121], [122], [124], [125], [126]. Further, radiation can induce the formation of brittle phases along grain boundaries and other defect sinks that can reduce ductility and degrade fracture toughness [124], [125], [127].
第一组（SFR、LFR、MSR、TWR）所面临的高温、高剂量运行环境挑战，将使强度、蠕变和蠕变疲劳行为的重要性进一步提升，同时低温下的断裂韧性也备受关注。尽管冷却剂与结构材料之间的化学相互作用会带来一些降解挑战，但主要担忧是核部件中预期的极高辐射损伤水平。在这种高损伤水平下，主要的降解模式很可能由空位肿胀和相稳定性驱动。空位肿胀发生在 0.35–0.55T M 的同义温度范围内，对于钢材（325–650 °C），这一温度范围与这四种高温、高剂量概念的反应堆核心温度范围重叠。图 16 比较了在快裂变反应堆谱中经 400–550 °C 辐照至高剂量后，Type 304L [116]、316 [12]以及 Ti 改性（D9）[12]奥氏体不锈钢和 9–12% Cr 回火铁素体/马氏体钢[12]、[117]、[118]、[119]的空位肿胀行为（通过浸泡密度测量获得）。 在所有情况下，空位肿胀行为包括初始的低肿胀瞬态阶段，随后是高肿胀率阶段（对于奥氏体钢接近 1% dpa[12]）。尽管结构应用中允许的最大体积肿胀量取决于设计，但基于典型的工程设计考虑，体积肿胀量>5%通常是不被接受的。当体积肿胀量>10%时，在辐照奥氏体钢中也观察到了严重的脆化现象[120]。多年对奥氏体合金的研究与开发通过利用肿胀抗性微观结构（如 Ti 改性奥氏体钢中的细小 TiC 析出物）成功延长了低肿胀瞬态阶段，但稳态空位增长的延迟不足以避免在运行到几代四核反应堆概念所考虑的高剂量期间出现显著的空位肿胀（参见图 9，图 10）。因此，正在考虑更多抗辐射合金，如 2.5–12% Cr 贝氏体-铁素体-马氏体钢，用于高剂量核心内部和反应堆压力容器应用。 加剧这一问题的是由于辐射诱导或增强的溶质偏析[121], [122]以及高能位移级联引起的沉淀物弹道溶解[122]导致的相不稳定。辐照可以成核或溶解相，改变基体的溶质成分并促进空洞生长[123]。此外，为提高合金强度而添加的粒子的溶解会导致软化，并损害高温强度和蠕变性能。例如，用于高温应用的镍基合金中提供强度的γ′基体沉淀物在辐照下不稳定[121], [122], [124], [125], [126]。此外，辐射可以在晶界和其他缺陷陷阱处诱导脆性相的形成，从而降低延展性并降低断裂韧性[124], [125], [127]。
第一组（SFR、LFR、MSR、TWR）所面临的高温、高剂量运行环境挑战，将使强度、蠕变和蠕变疲劳行为的重要性进一步提升，同时低温下的断裂韧性也备受关注。尽管冷却剂与结构材料之间的化学相互作用会带来一些降解挑战，但主要担忧是核部件中预期的极高辐射损伤水平。在这种高损伤水平下，主要的降解模式很可能由空位肿胀和相稳定性驱动。空位肿胀发生在 0.35–0.55T M 的同义温度范围内，对于钢材（325–650 °C），这一温度范围与这四种高温、高剂量概念的反应堆核心温度范围重叠。图 16 比较了在快裂变反应堆谱中经 400–550 °C 辐照至高剂量后，Type 304L [116]、316 [12]以及 Ti 改性（D9）[12]奥氏体不锈钢和 9–12% Cr 回火铁素体/马氏体钢[12]、[117]、[118]、[119]的空位肿胀行为（通过浸泡密度测量获得）。 在所有情况下，空位肿胀行为包括初始的低肿胀瞬态阶段，随后是高肿胀率阶段（对于奥氏体钢接近 1% dpa[12]）。尽管结构应用中允许的最大体积肿胀量取决于设计，但基于典型的工程设计考虑，体积肿胀量>5%通常是不被接受的。当体积肿胀量>10%时，在辐照奥氏体钢中也观察到了严重的脆化现象[120]。多年对奥氏体合金的研究与开发通过利用肿胀抗性微观结构（如 Ti 改性奥氏体钢中的细小 TiC 析出物）成功延长了低肿胀瞬态阶段，但稳态空位增长的延迟不足以避免在运行到几代四核反应堆概念所考虑的高剂量期间出现显著的空位肿胀（参见图 9，图 10）。因此，正在考虑更多抗辐射合金，如 2.5–12% Cr 贝氏体-铁素体-马氏体钢，用于高剂量核心内部和反应堆压力容器应用。 加剧这一问题的是由于辐射诱导或增强的溶质偏析[121], [122]以及高能位移级联引起的沉淀物弹道溶解[122]导致的相不稳定。辐照可以成核或溶解相，改变基体的溶质成分并促进空洞生长[123]。此外，为提高合金强度而添加的粒子的溶解会导致软化，并损害高温强度和蠕变性能。例如，用于高温应用的镍基合金中提供强度的γ′基体沉淀物在辐照下不稳定[121], [122], [124], [125], [126]。此外，辐射可以在晶界和其他缺陷陷阱处诱导脆性相的形成，从而降低延展性并降低断裂韧性[124], [125], [127]。
Fig. 16. Comparison of the volumetric void swelling behavior of Type 304L [116], 316 [12] and a Ti-modified (D9) [12] austenitic stainless steel and 9–12% Cr-tempered ferritic/martensitic steels [12], [117], [118], [119] following irradiation at ∼400–550 °C to high doses in a fast fission reactor spectrum.
图 16. 在快堆谱中经 400–550 °C 辐照至高剂量后，304L 型[116]、316 型[12]以及 Ti 改性（D9）[12]奥氏体不锈钢和 9–12% Cr 回火铁素体/马氏体钢[12]、[117]、[118]、[119]的体积孔隙肿胀行为比较。
However, some types of second-phase particles have demonstrated good stability to high-dose neutron irradiation and can provide strength at high temperature while ameliorating radiation damage effects. For example, controlled additions of Ti and P to austenitic Fe–Cr–Ni alloys has been demonstrated to produce fine dispersions of TiC or M2P (M = Fe, etc.) precipitates that provide dramatic improvement in void swelling resistance after high-dose (∼100 dpa) irradiation compared to standard Fe–Cr–Ni alloys [51], [128], [129], [130]. Similarly, the nanometer-sized (Y,Ti,O)-rich particles in oxide-dispersion-strengthened alloys appear to exhibit good stability under irradiation and provide significant strength advantages over ferritic–martensitic alloys up to high temperatures [131], [132], [133], [134], [135]. An added benefit of nanometer-sized oxides is their role in healing radiation damage. The high particle density presents a very large surface area for point defect trapping, promoting self-healing via recombination [56] and thus keeping the net accumulated radiation damage at a low value.
然而，某些类型的第二相粒子已显示出对高剂量中子辐照的良好稳定性，并在高温下提供强度，同时改善辐照损伤效应。例如，向奥氏体 Fe–Cr–Ni 合金中控制添加 Ti 和 P 已被证明可以产生 TiC 或 M 2 P（M = Fe 等）析出物的细小弥散，与标准 Fe–Cr–Ni 合金相比，在高剂量（~100 dpa）辐照后，空位肿胀抗力有显著提高[51], [128], [129], [130]。类似地，氧化物弥散强化合金中的纳米级(Y,Ti,O)富集颗粒似乎在辐照下表现出良好稳定性，并在高温下比铁素体-马氏体合金提供显著强度优势[131], [132], [133], [134], [135]。纳米级氧化物的额外好处在于它们在修复辐照损伤中的作用。高颗粒密度提供了非常大的表面积用于点缺陷捕获，通过复合促进自修复[56]，从而将净累积辐照损伤保持在较低水平。
然而，某些类型的第二相粒子已显示出对高剂量中子辐照的良好稳定性，并在高温下提供强度，同时改善辐照损伤效应。例如，向奥氏体 Fe–Cr–Ni 合金中控制添加 Ti 和 P 已被证明可以产生 TiC 或 M 2 P（M = Fe 等）析出物的细小弥散，与标准 Fe–Cr–Ni 合金相比，在高剂量（~100 dpa）辐照后，空位肿胀抗力有显著提高[51], [128], [129], [130]。类似地，氧化物弥散强化合金中的纳米级(Y,Ti,O)富集颗粒似乎在辐照下表现出良好稳定性，并在高温下比铁素体-马氏体合金提供显著强度优势[131], [132], [133], [134], [135]。纳米级氧化物的额外好处在于它们在修复辐照损伤中的作用。高颗粒密度提供了非常大的表面积用于点缺陷捕获，通过复合促进自修复[56]，从而将净累积辐照损伤保持在较低水平。
3.2.2.2. Very high-temperature gas-cooled concepts
3.2.2.2. 超高温气体冷却概念
The structural materials challenges become magnified considerably when moving from medium-temperature designs to very-high-temperature designs in which materials must withstand temperatures approaching 1000 °C (GFR, VHTR). Corrosion and oxidation of alloys are unavoidable at these temperatures due to the very rapid kinetics. In all cases, the challenge is to develop coolant–materials systems that result in the formation of protective and self-healing films to ensure the longevity of the structures for the life of the reactor.
当从中等温度设计转向非常高温设计时，结构材料面临的挑战会显著加剧，在这些设计中，材料必须承受接近 1000°C 的温度（GFR、VHTR）。由于动力学非常迅速，合金在这些温度下的腐蚀和氧化是不可避免的。在所有情况下，挑战都是开发冷却剂-材料系统，形成保护性和自愈性薄膜，以确保结构在反应堆寿命期间的长久性。
当从中等温度设计转向非常高温设计时，结构材料面临的挑战会显著加剧，在这些设计中，材料必须承受接近 1000°C 的温度（GFR、VHTR）。由于动力学非常迅速，合金在这些温度下的腐蚀和氧化是不可避免的。在所有情况下，挑战都是开发冷却剂-材料系统，形成保护性和自愈性薄膜，以确保结构在反应堆寿命期间的长久性。
At the extreme operating temperatures envisioned for gas-cooled reactor concepts, graphite and ceramic composites are the leading candidates for structural materials [136]. Along with the numerous engineering design issues associated with utilization of low-ductility materials in a complex high-power energy system, the property degradation associated with neutron displacement damage poses particular challenges. The anisotropic response of graphite to neutron displacement damage due to its hexagonal close-packed crystal structure requires the use of specially manufactured “nuclear” grades of graphite to achieve the desired component lifetimes [136]. For components that are subject to relatively high displacement damage exposures or engineering stresses, ceramic composites must be used instead of graphite. Nevertheless, metals are still the most viable materials for heat transfer components such as heat exchangers.
在气冷反应堆概念所设想的极端工作温度下，石墨和陶瓷复合材料是结构材料的领先候选者[136]。除了在复杂高功率能源系统中使用低延展性材料所带来的众多工程设计问题外，中子位移损伤引起的性能退化也构成了特殊挑战。由于石墨的六方密堆积晶体结构对其中子位移损伤的各向异性响应，需要使用专门制造的“核级”石墨才能实现所需的部件寿命[136]。对于承受相对较高位移损伤暴露或工程应力的部件，必须使用陶瓷复合材料代替石墨。尽管如此，金属仍然是换热器等传热部件最可行的材料。
在气冷反应堆概念所设想的极端工作温度下，石墨和陶瓷复合材料是结构材料的领先候选者[136]。除了在复杂高功率能源系统中使用低延展性材料所带来的众多工程设计问题外，中子位移损伤引起的性能退化也构成了特殊挑战。由于石墨的六方密堆积晶体结构对其中子位移损伤的各向异性响应，需要使用专门制造的“核级”石墨才能实现所需的部件寿命[136]。对于承受相对较高位移损伤暴露或工程应力的部件，必须使用陶瓷复合材料代替石墨。尽管如此，金属仍然是换热器等传热部件最可行的材料。
In this concept, high-temperature helium gas will pass through an intermediate heat exchanger, where it will transfer heat to a secondary coolant. Such temperatures require the use of nickel-based alloys rich in chromium (about 22 wt.%) and strengthened by additions of Mo, Co and W (for example, Inconel 617 and Haynes 230®) [137]. The helium inevitably contains parts per million (ppm) levels of CO, CO2, H2, H2O, and CH4 as impurities, which arise mainly from reactions between the hot graphite core and in-leakage of O2, N2 and water vapor from seals and welds, and degassing of reactor materials such as fuel, thermal insulation and in-core structural materials [138], [139].
在这个概念中，高温氦气将流经一个中间换热器，在那里它将热量传递给二级冷却剂。这样的温度需要使用富含铬（约 22 wt.%）并由 Mo、Co 和 W 添加物强化的镍基合金（例如 Inconel 617 和 Haynes 230®）[137]。氦气不可避免地含有百万分之几（ppm）水平的 CO、CO 2 、H 2 、H 2 O 和 CH 4 作为杂质，这些杂质主要来自高温石墨核心与 O 2 、N 2 和水蒸气的泄漏反应，以及燃料、隔热材料和核内结构材料等反应堆材料的脱气[138]、[139]。
在这个概念中，高温氦气将流经一个中间换热器，在那里它将热量传递给二级冷却剂。这样的温度需要使用富含铬（约 22 wt.%）并由 Mo、Co 和 W 添加物强化的镍基合金（例如 Inconel 617 和 Haynes 230®）[137]。氦气不可避免地含有百万分之几（ppm）水平的 CO、CO 2 、H 2 、H 2 O 和 CH 4 作为杂质，这些杂质主要来自高温石墨核心与 O 2 、N 2 和水蒸气的泄漏反应，以及燃料、隔热材料和核内结构材料等反应堆材料的脱气[138]、[139]。
Depending on the impurity concentration, temperature and alloy composition, the impurities react with the metallic surfaces of the heat exchanger resulting in oxidation, oxide reduction, carburization and decarburization. Chromium oxide is stable at oxygen partial pressures above a critical value and reduces at partial pressures below this value. Similarly, chromium carbide is stable above a critical carbon activity, and decarburization is expected to occur below the critical value. Oxidation, decarburization and carburization can degrade the mechanical properties of the alloy; for example, oxidation reduces the load-bearing cross-section of the component and internal oxide precipitates act as the preferential crack initiation sites [140] near the surface of the alloy, which can, potentially, decrease the creep and fatigue life of the alloy. A significant reduction in the creep-rupture ductility of alloy 800H [141], alloy 617 [142] and Hastelloy X [143] has been reported in a carburizing environment in comparison to pure helium and air environments; Fig. 17 shows an example of the effect of oxygen and methane impurities on the creep rupture behavior of alloy 617 (note the longer creep lifetime but lower tertiary creep regime for the methane impurity condition). A coarse and semi-continuous film of carbides forms along the grain boundaries during carburization, and this may act as preferential crack initiation and propagation paths, and could decrease the operating life of the alloys. Grain boundary migration and sliding has been identified as the dominant creep deformation mechanism in the candidate alloys, such as alloy 617 at 1000 °C [144], [145], and the dissolution of carbides due to decarburization may lead to significant loss of the creep strength. Therefore, a detailed knowledge of the oxidation mechanisms and rates of microstructure degradation is important to estimate the lifetime of the component and define mitigation strategies for improved oxidation performance of alloys. However, it is unlikely that unprotected alloys can maintain their integrity at 1000 °C without protection by coatings or barrier layers. Development of protective layers without compromising thermal conductivity is perhaps the most important major challenge for structural materials for the VHTR environment.
根据杂质浓度、温度和合金成分的不同，杂质会与换热器的金属表面发生反应，导致氧化、氧化物还原、渗碳和脱碳。氧化铬在氧分压高于临界值时稳定，在分压低于该值时会还原。类似地，碳化铬在碳活性高于临界值时稳定，在低于临界值时会发生脱碳。氧化、脱碳和渗碳会降低合金的力学性能；例如，氧化会减少构件的承载截面，内部氧化物析出物作为合金表面附近的优先裂纹起始点[140]，这可能会降低合金的蠕变和疲劳寿命。与纯氦气和空气环境相比，在渗碳环境中，合金 800H[141]、合金 617[142]和哈氏合金 X[143]的蠕变断裂延展性显著降低；图 17 展示了氧气和甲烷杂质对合金 617 蠕变断裂行为的影响（注意甲烷杂质条件下蠕变寿命更长但次级蠕变阶段更低）。在渗碳过程中，沿晶界形成粗大的半连续碳化物膜，这可能作为优先的裂纹萌生和扩展路径，并可能降低合金的使用寿命。晶界迁移和滑动已被确定为候选合金（如 1000°C 下的合金 617）的主要蠕变变形机制[144][145]，而脱碳导致的碳化物溶解可能导致蠕变强度显著下降。因此，详细了解微观结构退化的氧化机制和速率对于估算部件寿命和制定改善合金抗氧化性能的缓解策略至关重要。然而，如果没有涂层或屏障层的保护，未保护的合金不太可能保持在 1000°C 时保持完整性。 在 VHTR 环境下，开发不牺牲导热性的防护层，可能是结构材料面临的最重要主要挑战。
根据杂质浓度、温度和合金成分的不同，杂质会与换热器的金属表面发生反应，导致氧化、氧化物还原、渗碳和脱碳。氧化铬在氧分压高于临界值时稳定，在分压低于该值时会还原。类似地，碳化铬在碳活性高于临界值时稳定，在低于临界值时会发生脱碳。氧化、脱碳和渗碳会降低合金的力学性能；例如，氧化会减少构件的承载截面，内部氧化物析出物作为合金表面附近的优先裂纹起始点[140]，这可能会降低合金的蠕变和疲劳寿命。与纯氦气和空气环境相比，在渗碳环境中，合金 800H[141]、合金 617[142]和哈氏合金 X[143]的蠕变断裂延展性显著降低；图 17 展示了氧气和甲烷杂质对合金 617 蠕变断裂行为的影响（注意甲烷杂质条件下蠕变寿命更长但次级蠕变阶段更低）。在渗碳过程中，沿晶界形成粗大的半连续碳化物膜，这可能作为优先的裂纹萌生和扩展路径，并可能降低合金的使用寿命。晶界迁移和滑动已被确定为候选合金（如 1000°C 下的合金 617）的主要蠕变变形机制[144][145]，而脱碳导致的碳化物溶解可能导致蠕变强度显著下降。因此，详细了解微观结构退化的氧化机制和速率对于估算部件寿命和制定改善合金抗氧化性能的缓解策略至关重要。然而，如果没有涂层或屏障层的保护，未保护的合金不太可能保持在 1000°C 时保持完整性。 在 VHTR 环境下，开发不牺牲导热性的防护层，可能是结构材料面临的最重要主要挑战。
Fig. 17. Effect of methane and oxygen impurities in helium on the creep rupture behavior of Alloy 617 at 843 °C [142].
图 17. 氦气中甲烷和氧气杂质对合金 617 在 843°C 下的蠕变断裂行为的影响[142]。
3.2.2.3. The SCWR concept
3.2.2.3. SCWR 概念
The SCWR is unique in that it is the only water-cooled concept among the Generation IV designs. Maintaining water in the supercritical state requires very high pressures and exposes the entire circuit to water temperatures that are well above the existing knowledge base. Until recently, processes such as IGSCC and IASCC in supercritical water have been relatively unexplored. Recent research has shown that alloys that are susceptible to these degradation modes in LWR environments also show susceptibility in SCW, though at higher rates due to the higher temperatures [146]. Alloys considered for structural components fall broadly into two classes: austenitic stainless steels and nickel base alloys are resistant to general corrosion, but susceptible to IGSCC and IASCC, whereas ferritic–martensitic alloys are resistant to SCC but generally exhibit higher rates of oxidation. For iron-based systems, corrosion resistance is associated with the ability of iron- and chromium-based surface oxides to function as barriers to the transport of reactants (oxygen and metal ions). These oxides are the same as those typically formed in other oxidizing environments, such as steam, under less extreme conditions. The main issues related to the behavior of these oxides as protective layers in SCW are similar to those for advanced-steam-cycle fossil energy plants [147]. However, wall thicknesses of critical core components such as fuel cladding and coolant tubes are an order of magnitude smaller than boiler tubes in fossil plants, placing a greater burden on the development of thin protective oxide layers.
超临界水冷反应堆（SCWR）在第四代核能设计中是独一无二的，因为它是唯一采用水冷的概念。维持水处于超临界状态需要极高的压力，并将整个回路暴露在远超现有知识基础的水温中。直到最近，超临界水中的 IGSCC 和 IASCC 等过程相对未被充分探索。近期研究表明，在轻水反应堆环境中易受这些退化模式影响的合金，在超临界水中也表现出易感性，但由于温度更高，其退化速率更快[146]。用于结构组件的合金大致可分为两类：奥氏体不锈钢和镍基合金对一般腐蚀具有抵抗力，但对 IGSCC 和 IASCC 易感，而铁素体-马氏体合金对应力腐蚀开裂具有抵抗力，但通常表现出更高的氧化速率。对于铁基系统，耐腐蚀性与其铁基和铬基表面氧化物的能力有关，这些氧化物能够作为反应物（氧气和金属离子）传输的屏障。 这些氧化物与其他氧化环境（如蒸汽）在较温和条件下形成的氧化物相同。这些氧化物作为超临界水冷堆中的保护层的行为所面临的主要问题，与先进蒸汽循环的化石能源电厂相似[147]。然而，燃料包壳和冷却管等关键核心部件的壁厚比化石电厂的锅炉管小一个数量级，这给薄保护氧化物层的发展带来了更大的挑战。
超临界水冷反应堆（SCWR）在第四代核能设计中是独一无二的，因为它是唯一采用水冷的概念。维持水处于超临界状态需要极高的压力，并将整个回路暴露在远超现有知识基础的水温中。直到最近，超临界水中的 IGSCC 和 IASCC 等过程相对未被充分探索。近期研究表明，在轻水反应堆环境中易受这些退化模式影响的合金，在超临界水中也表现出易感性，但由于温度更高，其退化速率更快[146]。用于结构组件的合金大致可分为两类：奥氏体不锈钢和镍基合金对一般腐蚀具有抵抗力，但对 IGSCC 和 IASCC 易感，而铁素体-马氏体合金对应力腐蚀开裂具有抵抗力，但通常表现出更高的氧化速率。对于铁基系统，耐腐蚀性与其铁基和铬基表面氧化物的能力有关，这些氧化物能够作为反应物（氧气和金属离子）传输的屏障。 这些氧化物与其他氧化环境（如蒸汽）在较温和条件下形成的氧化物相同。这些氧化物作为超临界水冷堆中的保护层的行为所面临的主要问题，与先进蒸汽循环的化石能源电厂相似[147]。然而，燃料包壳和冷却管等关键核心部件的壁厚比化石电厂的锅炉管小一个数量级，这给薄保护氧化物层的发展带来了更大的挑战。
While oxide growth rates and product morphological details are specific to the oxygen content of the fluid, the temperature and the steel composition, and possibly other factors [148], the oxide structure on steels after exposure to SCW is similar to what is observed under steam conditions for ferritic and ferritic–martensitic (F-M) steels [147]. Because it is generally accepted that chromia-containing spinels are better permeation barriers to cations (metal) and anions (oxygen, OH, etc.) relative to iron oxides [149], it is the underlying oxide layer that can proffer the best corrosion resistance in the SCW environment. This has been observed in recent work on steels under nuclear SCW conditions [148], [150], [151] in terms of increasing corrosion resistance with increasing chromium content of the alloy.
氧化物的生长速率和产物形态细节取决于流体的氧含量、温度以及钢的成分，并可能受到其他因素的影响[148]，但在超临界水(SCW)环境中暴露后的钢中氧化物结构，与铁素体和铁素体-马氏体(F-M)钢在蒸汽条件下的观察结果相似[147]。由于普遍认为含铬尖晶石相对于氧化铁更能有效阻挡阳离子（金属）和阴离子（氧、OH 等）的渗透[149]，因此，在 SCW 环境中，基体氧化物层能够提供最佳的耐腐蚀性能。这一现象在最近关于核级 SCW 条件下钢的研究工作中得到证实[148]、[150]、[151]，即随着合金中铬含量的增加，耐腐蚀性能也随之提高。
氧化物的生长速率和产物形态细节取决于流体的氧含量、温度以及钢的成分，并可能受到其他因素的影响[148]，但在超临界水(SCW)环境中暴露后的钢中氧化物结构，与铁素体和铁素体-马氏体(F-M)钢在蒸汽条件下的观察结果相似[147]。由于普遍认为含铬尖晶石相对于氧化铁更能有效阻挡阳离子（金属）和阴离子（氧、OH 等）的渗透[149]，因此，在 SCW 环境中，基体氧化物层能够提供最佳的耐腐蚀性能。这一现象在最近关于核级 SCW 条件下钢的研究工作中得到证实[148]、[150]、[151]，即随着合金中铬含量的增加，耐腐蚀性能也随之提高。
The most daunting challenge for materials in the SCW environment is resistance to SCC and IASCC. While nickel-base alloys and austenitic stainless steels are very resistant to corrosion in SCW, they are most susceptible to SCC. Intergranular stress corrosion cracking occurs readily in high-purity, deaerated SCW at 400 °C and above in both austenitic stainless steels and nickel-base alloys. Cracking severity increases exponentially with temperature in both stainless steels and nickel-base alloys [152]. Over this same temperature range of 400–600 °C, ferritic–martensitic alloys are resistant to SCC [151], [153], [154]. Furthermore, SCC of susceptible alloys is known to be exacerbated by persistent radiation damage of the metal [155].
在 SCW 环境中，材料面临的最严峻挑战是抗应力腐蚀开裂（SCC）和抗晶间应力腐蚀开裂（IASCC）。虽然镍基合金和奥氏体不锈钢在 SCW 中的耐腐蚀性非常好，但它们最容易发生 SCC。在 400°C 及以上的高纯度脱气 SCW 中，奥氏体不锈钢和镍基合金都容易发生晶间应力腐蚀开裂。在奥氏体不锈钢和镍基合金中，裂纹的严重程度随着温度的升高呈指数级增加[152]。在同一 400–600°C 的温度范围内，铁素体-马氏体合金对 SCC 具有抗性[151], [153], [154]。此外，已知易感合金的 SCC 会因金属的持续辐射损伤而加剧[155]。
在 SCW 环境中，材料面临的最严峻挑战是抗应力腐蚀开裂（SCC）和抗晶间应力腐蚀开裂（IASCC）。虽然镍基合金和奥氏体不锈钢在 SCW 中的耐腐蚀性非常好，但它们最容易发生 SCC。在 400°C 及以上的高纯度脱气 SCW 中，奥氏体不锈钢和镍基合金都容易发生晶间应力腐蚀开裂。在奥氏体不锈钢和镍基合金中，裂纹的严重程度随着温度的升高呈指数级增加[152]。在同一 400–600°C 的温度范围内，铁素体-马氏体合金对 SCC 具有抗性[151], [153], [154]。此外，已知易感合金的 SCC 会因金属的持续辐射损伤而加剧[155]。
Radiation effects on IGSCC are only now being investigated for SCW conditions, yet results show that irradiation significantly increases the extent of SCC in stainless steels and nickel-base alloys. Proton irradiations of type 316L stainless steel and Ni-based alloy 690 showed a significant increase in intergranular cracking relative to the unirradiated cases. The increased cracking could not be attributed to radiation-induced segregation or hardening alone, so combinations of factors or other defect mechanisms must be at play [156]. Both the density of cracks and the crack depth increased over the unirradiated case following irradiation to 7 dpa and testing in SCW at 400 °C. One set of data exists on the effect of neutron irradiation on cracking in SCW, where an austenitic stainless steel was irradiated to doses of over 40 dpa and showed extreme embrittlement [157]. Under the same irradiation and testing conditions, ferritic–martensitic alloys were found to be resistant to cracking.
辐照对 IGSCC 的影响目前仅在 SCW 条件下进行研究，但结果表明辐照会显著增加不锈钢和镍基合金的应力腐蚀开裂程度。316L 不锈钢和镍基合金 690 的质子辐照实验显示，与未辐照样品相比，晶间开裂显著增加。这种开裂的增加不能单独归因于辐照引起的偏析或硬化，因此必然存在多种因素组合或其他缺陷机制在起作用[156]。经过 7 dpa 辐照并在 400 °C 的 SCW 条件下测试后，裂纹密度和裂纹深度均比未辐照情况有所增加。目前有一组关于中子辐照对 SCW 条件下开裂影响的数据，其中一种奥氏体不锈钢辐照剂量超过 40 dpa 后表现出极端脆化[157]。在相同的辐照和测试条件下，铁素体-马氏体合金则表现出抗开裂性能。
辐照对 IGSCC 的影响目前仅在 SCW 条件下进行研究，但结果表明辐照会显著增加不锈钢和镍基合金的应力腐蚀开裂程度。316L 不锈钢和镍基合金 690 的质子辐照实验显示，与未辐照样品相比，晶间开裂显著增加。这种开裂的增加不能单独归因于辐照引起的偏析或硬化，因此必然存在多种因素组合或其他缺陷机制在起作用[156]。经过 7 dpa 辐照并在 400 °C 的 SCW 条件下测试后，裂纹密度和裂纹深度均比未辐照情况有所增加。目前有一组关于中子辐照对 SCW 条件下开裂影响的数据，其中一种奥氏体不锈钢辐照剂量超过 40 dpa 后表现出极端脆化[157]。在相同的辐照和测试条件下，铁素体-马氏体合金则表现出抗开裂性能。
4. Conclusions 4. 结论
The continued utilization of nuclear energy systems for worldwide baseload electricity offers a number of materials research challenges. The high reliability of current light-water fission reactors (e.g. 90% average capacity factor by US reactors for the past decade) demonstrates the high reliability of this energy source under normal operating conditions. Planned extensions in the operating lifetime for reactors are being supported by accompanying materials R&D to investigate corrosion and neutron-induced materials degradation phenomena. The three major materials challenges for continued safe, reliable and cost-effective utilization of water-cooled nuclear reactors for electricity production are development of improved understanding of the synergistic fundamental mechanisms responsible for corrosion and stress corrosion cracking degradation of austenitic steels and nickel base alloys, development of a truly predictive understanding of the multi-physics phenomena responsible for radiation hardening and degradation in ductility and fracture toughness of complex structural alloys (in particular RPV steels), and nuclear fuels innovations including investigation of further improvements in the reliability of LWR fuel systems (can an additional order of magnitude improvement be achieved on top of the three orders of magnitude improvement accomplished over the past 40 years?), as well as exploration of new LWR fuel systems with improved accident tolerance without reducing the favorable performance and reliability features achieved by current fuels under normal operating conditions. The eventual development of advanced Generation IV fission reactor systems is directly linked to successful resolution of several daunting materials challenges associated with the higher radiation damage levels and/or operating temperatures for these concepts and aggressive coolants, all of which are detrimental to the reliability of structural materials. The materials research challenges that need to be successfully resolved for fusion energy systems are even more daunting due to the high concentrations of H and He gases that will be produced in the materials as a result of the high-energy fusion neutron spectrum. A significant practical issue that is a barrier to the development of structural materials for Generation IV fission and fusion energy concepts is the limited worldwide capability for high-flux materials irradiations; there are no high-intensity fusion neutron irradiation facilities, and the number of fast fission test reactors is dwindling.
继续利用核能系统为全球基础电力提供支持，带来了一系列材料研究挑战。当前轻水裂变反应堆（例如，美国反应堆过去十年的平均容量因子为 90%）的高可靠性，证明了这种能源在正常操作条件下的高可靠性。为延长反应堆运行寿命所做的计划，正得到伴随材料研发的支持，以研究腐蚀和中子诱导的材料退化现象。 为持续安全、可靠且经济高效地利用水冷核反应堆进行电力生产，面临三大主要材料挑战：一是深化对奥氏体钢和镍基合金腐蚀及应力腐蚀开裂降解的协同基础机理的理解；二是发展真正可预测的、关于多物理现象的理解，这些现象导致复杂结构合金（尤其是反应堆压力容器钢）的延展性和断裂韧性的辐照硬化与降解；三是核燃料创新，包括研究进一步提高轻水反应堆燃料系统可靠性的方法（是否能在过去 40 年取得的三个数量级改进基础上再提升一个数量级？），以及探索具有更高事故耐受性的新型轻水反应堆燃料系统，同时不降低当前燃料在正常操作条件下的优良性能和可靠性特征。 先进第四代裂变反应堆系统的最终发展直接与成功解决这些概念和激进冷却剂所面临的严峻材料挑战相关，这些挑战包括更高的辐射损伤水平和/或工作温度，所有这些都会损害结构材料的可靠性。由于高能聚变中子谱导致材料中产生高浓度的氢和氦气体，因此需要成功解决聚变能源系统所需材料研究挑战，其难度更大。阻碍第四代裂变和聚变能源概念结构材料发展的一个重要实际问题是对高通量材料辐照的全球能力有限；没有高强度的聚变中子辐照设施，而且快堆试验反应堆的数量正在减少。
继续利用核能系统为全球基础电力提供支持，带来了一系列材料研究挑战。当前轻水裂变反应堆（例如，美国反应堆过去十年的平均容量因子为 90%）的高可靠性，证明了这种能源在正常操作条件下的高可靠性。为延长反应堆运行寿命所做的计划，正得到伴随材料研发的支持，以研究腐蚀和中子诱导的材料退化现象。 为持续安全、可靠且经济高效地利用水冷核反应堆进行电力生产，面临三大主要材料挑战：一是深化对奥氏体钢和镍基合金腐蚀及应力腐蚀开裂降解的协同基础机理的理解；二是发展真正可预测的、关于多物理现象的理解，这些现象导致复杂结构合金（尤其是反应堆压力容器钢）的延展性和断裂韧性的辐照硬化与降解；三是核燃料创新，包括研究进一步提高轻水反应堆燃料系统可靠性的方法（是否能在过去 40 年取得的三个数量级改进基础上再提升一个数量级？），以及探索具有更高事故耐受性的新型轻水反应堆燃料系统，同时不降低当前燃料在正常操作条件下的优良性能和可靠性特征。 先进第四代裂变反应堆系统的最终发展直接与成功解决这些概念和激进冷却剂所面临的严峻材料挑战相关，这些挑战包括更高的辐射损伤水平和/或工作温度，所有这些都会损害结构材料的可靠性。由于高能聚变中子谱导致材料中产生高浓度的氢和氦气体，因此需要成功解决聚变能源系统所需材料研究挑战，其难度更大。阻碍第四代裂变和聚变能源概念结构材料发展的一个重要实际问题是对高通量材料辐照的全球能力有限；没有高强度的聚变中子辐照设施，而且快堆试验反应堆的数量正在减少。
Acknowledgements 致谢
The authors would like to acknowledge Roger Staehle for graciously providing several figures, and Jeremy Busby for comments on the draft manuscript. We also thank Robin Grimes (Imperial College) for helpful comments on fuels and materials for nuclear reactors, and Bo Cheng (Electric Power Research Institute), Kurt Terrani and Lance Snead (ORNL), and Kemal Pasamehmetoglu (Idaho National Laboratory) for input on considerations for accident tolerant fuels.
作者们感谢 Roger Staehle 慷慨提供数个插图，以及 Jeremy Busby 对初稿的评论。我们还感谢 Robin Grimes（帝国理工学院）对核反应堆燃料和材料的建设性意见，以及 Bo Cheng（电力研究院）、Kurt Terrani 和 Lance Snead（橡树岭国家实验室）以及 Kemal Pasamehmetoglu（爱达荷国家实验室）对事故耐受性燃料考虑因素的贡献。
作者们感谢 Roger Staehle 慷慨提供数个插图，以及 Jeremy Busby 对初稿的评论。我们还感谢 Robin Grimes（帝国理工学院）对核反应堆燃料和材料的建设性意见，以及 Bo Cheng（电力研究院）、Kurt Terrani 和 Lance Snead（橡树岭国家实验室）以及 Kemal Pasamehmetoglu（爱达荷国家实验室）对事故耐受性燃料考虑因素的贡献。
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