基于人工神经网络的RPV材料辐照脆化预测模型研究

在分析一定量随站测试样品的基础上，构建了具有较高精度的反应堆压力容器（RPV）材料韧脆转变温度（DBTT）预测的人工神经网络模型，并利用模型研究了中子注量和中子注量率对RPV材料DBTT的影响。结果表明，材料DBTT随着中子注量增加出现先线性上升，然后平缓上升，最后饱和的趋势，而中子注量率对RPV材料辐照脆化的影响不明显。      

疲劳强度减弱系数与应力集中系数在螺纹疲劳分析中的应用研究

以反应堆压力容器（RPV）为例,给出主螺栓螺纹的疲劳强度减弱系数（K_f）与应力集中系数（K_t）之间的理论关系、K_f与K_t的合理取值范围、工程上的等效判定方法以及在疲劳分析中的使用方法,并采用特定K_f值给出疲劳分析算例。明确了两者的概念及区别,在高强度钢材料螺纹结构的应用方面给出意见及范例,可为工程上螺栓选取及结构设计提供借鉴。      

高温气冷堆压力容器侧向支承地震易损性的敏感性分析

反应堆压力容器（RPV）侧向支承是高温气冷堆地震风险的关键贡献物项，对于反应堆地震安全至关重要。本文确定了高温气冷堆RPV侧向支承的地震易损性变量，分析出易损性变量因子的合理取值，计算得到侧向支承的地震易损性曲线和高置信度低失效概率（HCLPF）抗震能力，挑选出易损性变量中的关键参数，并研究了RPV侧向支承HCLPF抗震能力对易损性关键参数的敏感性。结果表明，侧向支承的抗震能力明显高于设计基准地震动，易损性对于关键参数变异并不敏感。      

压力容器-保温层流道变形条件下临界热流密度试验研究

针对压力容器外部冷却（ERVC）应用中的压力容器-保温层流道（RPV-保温层流道）变形问题，利用提高临界热通量影响因素（FIMR）的试验装置，在相同流量范围开展了变形条件下壁面临界热流密度（CHF）的试验研究，分析了流道变形和流量变化对压力容器（RPV）下封头壁面CHF的影响规律，获得了流道变形情况下ERVC的安全裕度。结果表明:随着RPV下封头角度升高，循环流量增加，下封头壁面CHF增大；与原型流道相比，变形流道下封头壁面CHF的变化幅度小于7%，流道变化的影响并不显著；变形流道中，下封头壁面安全裕量最小的位置与原型流道相同，其安全裕量略有提高。      

国产反应堆压力容器用16MND5钢的蠕变损伤本构模型研究

为了获得反应堆压力容器（RPV）材料在高温下的蠕变行为,保证RPV在严重事故工况下的完整性,本研究对国产RPV用16MND5钢的高温蠕变性能进行了测试,获得了600~900℃下材料的蠕变性能,并基于应变强化的基本蠕变本构模型与基于延性耗竭理论的蠕变损伤模型,建立了适用于16MND5钢的蠕变损伤本构模型,给出了材料的蠕变损伤模型参数。结果表明,本文提出的蠕变损伤本构模型的有限元模拟数据与试验数据符合性较好,验证了此蠕变损伤模型的正确性。该方法可用于严重事故情况下RPV的蠕变损伤分析,为RPV的完整性分析提供支持。      

反应堆压力容器低合金钢锻件超声检测时机探讨

为了确定反应堆压力容器(RPV)钢锻件最佳超声检测时机,本文采用分析的方法明确了RCC-M规范要求是对RPV低合金钢锻件在最终精加工后才进行超声检测。通过研究超声检测机理并结合其他标准的规定,综合对此时机的合理性进行了探讨,提出了更为合理的检测时机应为对RPV低合金钢锻件性能热处理之后、精加工前进行。     

三层熔池结构对AP1000反应堆压力容器外壁面热流密度的影响

严重事故后期，反应堆压力容器（RPV）下腔室内熔融物（U-Zr-O-Fe）可能发生分层。但目前采用的三层熔池结构分析模型之间有较大差异。本文采用了3种不同熔池结构模型:Esmaili & Khatib-Rahbar模型、Seiler模型、MAAP5程序模型分别计算了AP1000电厂的三层熔池结构并对RPV外壁面热流密度分布进行分析。结果表明，3种计算模型计算的熔池结构差异很大，进而影响了RPV外侧的热流密度分布。相比Esmaili & Khatib-Rahbar模型，Seiler模型更为保守。而MAAP5程序模型虽然在计算氧化物层和重金属层成分时是基于热力学理论，但轻金属层成分的确定方法尚待进一步改进。      

~（85）Sr、~（134）Cs和~（60）Co在浅地层黄土与地下水间的分配系

本文报道了￣（８５）Ｓｒ、￣（１３４）Ｃｓ和￣（６０）Ｃｏ在采自本院野外试验场，从地表到地下２ｍ深黄土与地下水间分配系数的测定结果，以及核素浓度、ｐＨ和温度对分配系数的影响。黄土采集的深度范围分别为：地表－０．５ｍ，０．５－１ｍ，１－１．５ｍ，１．５－２ｍ，共采集１６个样品；地下水取自试验场的井水。实验采用静态法。实验结果表明，１６个样品测定的￣（８５）Ｓｒ、￣（１３４）Ｃｓ和￣（６０）Ｃｏ的分配系数的均值分别为８２、７．４×１０￣３和５．４×１０￣３ｍＬ／ｇ；在不同深度的黄土中测得的分配系数基本一致。ｐＨ值的变化对￣（１３４）Ｃｓ和￣（８５）Ｓｒ的分配系数有较大影响，对￣（６０）Ｃｏ则影响不大。     

裂片核素在岩石中的迁移研究Ⅲ．纵向弥散系数的测定和核素~（129）Ⅰ的迁移模型

研究了￣（１２５）Ⅰ（模拟裂片核素￣（１２９）Ⅰ）在花岗岩中的吸附动力学，结果表明在实验条件下￣（１２５）Ⅰ吸附极少。由测得的渗透曲线和流量曲线经物料衡算，求得岩石输出浓度与时间的关系曲线。利用分布函数性质，求得纵向弥散系数Ｄ、孔隙流速ｖ和有效孔隙率ε＿ｅ。提出了￣（１２９）Ⅰ在花岗岩中迁移的数学模型方程，并求解得到核素浓度随时空变化的分析解和数值解。     

~（nat）W（p，xn）~（181─186）Re反应的激发函数

使用叠靶技术测量了２２．５ＭｅＶ的质子照射天然钨时生成~（１８１）Ｒｅ、~（１８２）Ｒｅ~ｍ、~（１８２）Ｒｅ~ｇ、~（１８３）Ｒｅ、~（１８４）Ｒｅ~ｍ、~（１８４）Ｒｅ~ｇ和~（１８６）Ｒｅ的激发函数。估算了~（１８６）Ｗ（ｐ，ｎ）~（１８６）Ｒｅ反应生产用于治疗的放射性核素~（１８６）Ｒｅ的厚靶产额，并与~（１８６）Ｗ（ｄ，２ｎ）~（１８６）Ｒｅ反应的有关数据进行了比较。     

评定结构材料临界脆性温度的新方法简介

在重要零部件（特别是核容器）的产品质量控制和防脆断安全分析中，经常使用以不同试验方法所取得的材料脆性断裂抗力参量，如冲击转变温度Ｔｃ和ＦＡＴＴ，无延性转变温度ＴＮＤＴ以及参考转变温度ＲＴＮＤＴ，近年来，原苏联学者在ＣＮＩＴＴ－ＭＡＳＣＨ开发了一种计算材料临界脆性温度的新方法，本文特就此作一简介。     

百万千瓦级核电站蒸汽发生器大锻件工艺评定

文章介绍了百万千瓦级核电站蒸汽发生器大锻件工艺评定的背景、依据、目的、技术指标和评定方法。根据蒸汽发生器锻件的结构特点和制造工艺,形成了一整套评定试验方案,可以对锻件各部位的化学成分、力学性能、金相组织和内部质量进行全面验证。该方案已在国内福清、方家山等多个核电工程中的蒸汽发生器锻件评定中成功应用。     

Manufacturing of ultra-large diameter 20 MnMoNi 5 5 steel forgings for reactor pressure vessels and their properties

As the structural material for RPV typical of increased dimensions, a set of ultra-large diameter steel forgings for a PHWR RPV is presented as outlined below.

1. (1) Material designation: 20 MnMoNi 5 5 (similar to SA508, Cl.3)

2. (2) Size of the forgings: flanges, 8.440 mm OD, a weight of 238 tons for shell flange; shells and torus, 7,920 mm OD, with large height; cover dome, 6,800 mm OD in chord and 460 mm thick; blank before formed to dome is ca. 8,000 mm OD.

3. (3) Chemical composition: particular effort was made for minimizing the tramp elements as P, S, As, Sn, Sb, Cu.

4. (4) Manufacturing, key points: steel making - combined refining and degassing in ladle; ingot making - largest size ingots, including 570 ton and 500 ton ingots; forging - special “outside-the-press” forging and forming techniques; heat treatment - prevention of H₂ flaking in normalizing and tempering and handling of the extra-large forgings at water quenching.

5. (5) Metallurgical properties: sufficiently uniform carbon distributions in the forgings; a lowest possible content of hydrogen, non-metallic inclusions and oxygen.

Mechanical properties: uniformity in tensile and toughness properties; flaws - only limited number of spots of UT indications under 2 mm EFG (EFS).     

A modified area function in time scale for the transmission data analysis of neutron resonance

A modified area function in time scale has been constructed with the phosphorescent decay of the detector (long-optical modes) and the background fraction included. A coefficient in the modified area function called self-shielding parameter is established to indicate the self-shielding effect in the experiment. The modified area function is independent from the equipment resolution. For the thick sample (nσ₀ ～ 10), the self-shielding effect would greatly improve the sensitivity of the modified area function to the temperature. Theoretically, we discuss the properties of the modified area function and estimate the least temperature deviation for this area method in the temperature measurement. Procedure of temperature measurement from isolated or multiple resonances by using the modified area method is also suggested.     

AP1000核岛主设备锻件性能提高对制造的影响

为了满足AP1000核电站设计寿命60年的需求,核岛设备蒸汽发生器锻件的强度和韧性要求比CPR1000核岛主设备都有所提高,加之尺寸增大,使得AP1000蒸汽发生器锻件的制造难度加大,对其变化认识不够,不仅锻件的产品质量不稳定,而且后序的焊接也可能出现质量问题,文章通过对比分析AP1000核电蒸汽发生器锻件与CPR1000锻件的变化,提出了采取措施的方向。     

Chromium-molybdenum steels for fusion reactor first walls — a review

Chromium-molybdenum steels have recently become of interest as a first wall and blanket structural material for fusion reactors. This application will be assessed, and possible approaches on how Cr---Mo alloys may be further developed for this application will be proposed.Generally, the Cr---Mo steels can be divided into two categories: unmodified, basically Cr---Mo---C steels and Cr---Mo---C steels modified by the addition of carbide-forming elements. Extensive research and development efforts have been conducted on the unmodified steels, especially Cr-1 Mo and 12 Cr---Mo steels. Considerable work has also been done on 12 Cr---Mo steels modified with additions of V, Nb, Ti and W. In recent years much of the research effort on this type of alloy has been directed at developing modified Cr---Mo steels with less than 12% Cr (generally 9%) for applications where the “stainless” properties imparted by chromium additions of at least 12% are not needed.We will examine the unmodified and modified steels in terms of hardenability, precipitation processes (stability at elevated temperatures), strength, and toughness. Where possible, we will discuss the effects of irradiation on these properties. Such a study leads to the types of tradeoffs that may be necessary between the well-researched unmodified Cr-1 Mo steel and a high-chromium modified steel.     

Current steel forgings and their properties for steam generator of nuclear power plant

Current steel forgings used for the construction of steam generators (SGs) for a nuclear power plant (NPP) were introduced from the following three (3) features. (1) Integral type steel forgings such as (i) primary head integrated with nozzles, manways and supports, (ii) steam drum head integrated with nozzle and handhole, (iii) conical shell integrated with cylindrical sections and handholes, have been developed to enhance the structural integrity of the component and to make fabrication and inspection including in-service inspection easier. (2) The high strength steel such as SA508 Cl.3a has been adopted to decrease the weight of components, resulting in enhancement of aseismatic properties of SG. (3) Low Si SA508 Cl.3 steel with the addition of Al was investigated and was applied to a heavy thick tube sheet forging, to minimize the macro- and micro-segregations in the ingot.     

非能动余热排出换热器冷凝换热性能研究

以浸没在高位水箱中的竖直管束为研究对象,对不同热负荷条件下竖直管束内冷凝换热特性进行研究,通过对比中心管与周围旁管的凝液增长率以及冷凝换热热阻,分析了中心管与旁管换热特性的差异,解释了低压条件下冷凝换热系数剧烈下降现象。将管束冷凝试验数据与已有单管试验数据对比发现,在相同蒸汽工况下,单管的冷凝换热系数与旁管的冷凝换热系数吻合较好,但远低于中心管的冷凝换热系数,说明中心管的换热性能相对于旁管确实得到了强化。通过对比换热系数的试验值与经典努塞尔理论和努塞尔修正理论的计算值发现,中心管的试验值与努塞尔修正理论计算值吻合较好,但旁管的偏差较大。     

压水堆核电厂核岛主设备国产化关注点——大型锻件

以核岛RPV和SG用低合金钢大型锻件为对象，分析了其重要性、供需、成熟性，以及法国20世纪七八十年代规模化发展核电的经验，力挺以进一步深入科研、集中、标准化的做法，加速提升国产大型锻件的供货能力。     

AP1000蒸汽发生器锻件制造的难点

AP1000蒸汽发生器是核岛关键设备,其制造要求严格。文章简要介绍了蒸汽发生器主体结构和材料,重点介绍了管板、锥形筒体和水室封头锻件的制造难点,并通过三门核电站1号机组和海阳核电站1号机组蒸汽发生器锻件的制造实践,指出在制造过程中需要引起关注的地方,总结了一些制造过程中发生的问题,提出了AP1000蒸汽发生器锻件国产化过程中的建议性措施。