AP1000核岛主设备用SA508-3钢锻件的监造

ASME SA508-3钢具有优越的可焊性、较好的抗中子辐照脆化性能和非常好的断裂韧性以及冲击韧性,因此被广泛应用于压水堆核电站核岛压力容器的制造中。AP1000三代核电机组的一些主设备,如反应堆压力容器、蒸汽发生器、稳压器的全部大锻件及一些重要部件均采用了这一钢种。通过对SA508-3钢锻件制造过程中的技术要点的分析,指出了该钢种的锻件在制造过程中的质量关注重点,提出了对该钢种锻件实施监造过程中的监督方法和监督重点。     

Current forging technology for integrated type steel forgings for nuclear steam supply system components

This brief review concerns the trend in using large size, integrated type steel forgings for nuclear steam supply system components as intended for easier in-service inspection (ISI).To achieve this trend, unique forging techniques have been developed.The forging techniques corresponding to each integrated type steel forging product together with a brief explanation of the development of forging facilities as well as a future aspects are introduced.     

真实表面材料及其老化效应对反应堆压力容器ERVC-CHF影响的试验研究

通过反应堆压力容器外部冷却（ERVC）实现熔融物堆内滞留（IVR）技术是核电厂严重事故缓解的重要措施之一。在本文的研究中,建立了二维切片式、全尺寸的试验台架FIRM,开展严重事故条件下反应堆压力容器ERVC-临界热流密度（CHF）试验研究。试验采用去离子水作为试验工质,获得了反应堆压力容器下封头ERVC过程的CHF限值。研究了真实表面材料对CHF的影响及其影响机理,讨论了在去离子水下表面材料SA508 Gr3. Cl.1钢的老化效应。本试验研究对于认识反应堆压力容器IVR-ERVC条件下的CHF行为、提高反应堆压力容器安全性有重要意义。     

Application of a new processing procedure for the further improvement of nuclear grade structural VCD steels

The enhancement of the quality of NiCrMoV-steels for turbine rotors and generator shafts by use of the vacuum carbon deoxidation (VCD) method is well known and established.So far the application of the VCD method has been limited to those steels which were not required to be made as fine grained steels and which consequently could be produced without the addition of aluminium.Klöckner Works have developed a method which allows the use of VCD — i.e. deoxidation of the steel by gaseous CO-reaction — with subsequent addition of Al for grain refinement. This new steel making process combines the advantages of a VCD steel with those of a fine grained steel.Seven production ingots of the steel 20 MnMoNi 5 5 have been produced by this process. The properties of the forgings made of these ingots are compared with forgings of Si/Al-deoxidized ingots. The VCD method has a beneficial influence of A-segregations. Sulphur prints show practically no segregation streaks. An improvement of impact properties and a greater isotropy of properties is gained by the VCD method.The excellent weldability of this steel could be demonstrated by weld simulation tests. There was be no susceptibility to cold cracking which was observed with SA 508 C1.3 forgings when the austenitic overlay cladding was welded on areas with A-segregations.In general the application of the VCD process to SA 508 C1.3 results in a remarkable improvement of material properties such as toughness and weldability and therefore renders components of higher reliability.     

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).     

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。     

氧化和工质中添加剂对SA508钢CHF影响的实验研究

在核电事故中当堆芯熔融物落入反应堆压力容器（RPV）下封头时,如果实际热流密度超过RPV的临界热流密度（CHF）,RPV将会被熔穿,造成事故的进一步扩大。为研究RPV在氧化条件下和有添加剂的工质中的CHF特性,采用池沸腾实验方法,以去离子水为工质,研究了RPV常用材料SA508钢经高温预氧化、7次池沸腾传热实验氧化后的CHF特性以及工质中添加剂对其CHF的影响。结果表明：在625 ℃下预氧化8 h后,SA508钢表面产生的较薄氧化层能增加传热面积、表面粗糙度和亲水性,从而提高CHF;随着池沸腾实验次数的增加,SA508钢表面的氧化腐蚀和颗粒沉积程度增加,CHF先增加后降低;0.4%硼酸（BA）、0.5%磷酸三钠（TSP）溶液和两者的混合溶液均有利于CHF的提升,但强化机理有所不同：BA会加速SA508钢表面的腐蚀并改善亲水性;TSP可降低表面张力使表面获得超亲水性;BA和TSP的混合溶液会形成一层沉积物使表面获得超亲水性。     

Application of positron annihilation line-shape analysis to fatigue damage for nuclear plant materials

Positron annihilation line-shape analysis is sufficiently sensitive to detect microstructural defects such as vacancies and dislocations. We are developing a portable positron annihilation system and applying this technique to fatigue damage in type 316 stainless steel and SA508 low alloy steel. The positron annihilation technique was found to be sensitive in the early fatigue life, i.e. up to 10% of the fatigue life, but showed little sensitivity in later stages of the fatigue life in type 316 stainless steel and SA508 low alloy steel. Type 316 stainless steel a higher positron annihilation sensitivity than that of SA508. It was considered that the amount of dislocation density change in the stainless steel was greater than that in the low alloy steel, because the initial microstructure contained a low dislocation density because of the solution heat treatment for the type 316 stainless steel.     

Effect of heat treatment and precipitation state on toughness of heavy section Mn-Mo-Ni-steel for nuclear power plants components

With regard to the mechanical properties of heavy forgings, made from steel grade 508 ASTM Class 3 (20 MnMoNi 5 5) the influence of the cooling intensity during hardening was investigated and the possibilities and limits of accelerating the cooling rate were studied. The importance of the transformation heat is discussed. It can be concluded from this, that it is impossible to avoid the formation of upper bainite as a typical microstructure of heavy forgings from this steel type.To obtain high impact energy the annealing condition must be conform with this type of microstructure. The reason for impact decrease by over-annealing is related to Mo₂C precipitation in the bainitic ferrite.Finally, experiences with intercritical heat treatment of this steel type are described and the special behaviour of the microstructure obtained is demonstrated.     

Application of high strength MnMoNi steel to pressure vessels for nuclear power plant

Recent increase in output of nuclear power plant has been attained by enlargement of major components such as pressure vessels. Such large components have almost reached a size limit from the points of manufacturing capacity and cost in both forgemasters and fabricaters. In order to solve this problem, it must be beneficial to apply design by use of material of higher strength, which brings reduction of pressure vessel thickness and weight. The Japan Steel Works Ltd. (JSW) has many manufacturing experiences of large integrated forgings made from high strength MnMoNi steel with tensile strength level of 620 MPa for steam generator (SG) pressure vessel, and has performed confirmation tests of its material properties. This paper describes the confirmation test results such as tensile and impact properties, nil-ductility transition temperature (NDT-T), static and dynamic fracture toughness, weldability including under-clad cracking (UCC) sensitivity, as well as metallurgical factors which influence on such material properties.     

Correlation of the thermodynamic calculation and the experimental observation of Ni-Mo-Cr low alloy steel changing Ni, Mo, and Cr contents

SA508 Gr.4N Ni-Mo-Cr low alloy steel has improved fracture toughness and strength compared to commercial low alloy steels such as SA508 Gr.3 Mn-Mo-Ni low alloy steel, which has less than 1% Ni. Higher strength and fracture toughness of low alloy steels can be achieved by increasing the Ni and Cr contents. In this study, the effects of the alloying elements of Ni and Cr on the microstructural characteristics and mechanical properties of SA508 Gr.4N Ni-Mo-Cr low alloy steel are evaluated. Changes in the stable phases of SA508 Gr.4N low alloy steel with these alloying elements were evaluated using thermodynamic calculation software. These values were then compared with the observed microstructural results. Additionally, tensile tests and Charpy impact test were carried out to evaluate the mechanical properties. The thermodynamic calculations show that Ni mainly affects the change of the matrix phase of γ and α rather than the carbide phase. Contrary to the Ni effect, Cr and Mo primarily affect the precipitation behavior of the carbide phases of Cr₂₃C₆, Cr₇C₃ and Mo₂C. In the microscopic observations, the lath martensitic structure becomes finer as the Ni content increases without affecting the carbides. When the Cr content decreases, the Cr carbide becomes unstable and carbide coarsening occurs. Carbide Mo₂C in the form of fine needles were observed in the high-Mo alloy. Greater strength was obtained after additions of Ni and Mo and the transition properties were improved as the Ni and Cr contents increased. These results were correlated with the thermodynamic calculation results.     

Experimental Research on Critical Heat Flux for Downward-facing Pool Boiling with SA508 Steel Plate

The external reactor vessel cooling (ERVC) is one of the important methods to achieve the in-vessel retention (IVR), while the critical heat flux (CHF) on the outside wall of the reactor pressure vessel (RPV) decides the maximum heat removal capacity of ERVC. In present work, a small CHF test facility was established. The test surface was made of SA508 steel which was the same surface material of prototype RPV. The deionized water was used as coolant in downward-facing CHF test under pool boiling condition. The influence of the real RPV material surface at different inclination angles and sub-cooling conditions on the CHF characteristics was studied. The influence of aging on CHF was also studied. The results show that the SA508 steel surface is easily oxidized, so its CHF is higher than that of copper and stainless steel surfaces. The CHF of SA508 steel surface increases with inclination angle, but there is a turning point near 30° and the CHF below the turning angle has no obvious trend with the increase of inclination angle. The CHF increases with the sub-cooling, and it shows linear growth characteristics. The test results provide a further understanding of the CHF behavior on the RPV outside wall and lay the foundation for future research work on CHF enhancement methods.     

池沸腾下朝向SA508钢表面临界热流密度特性试验研究

反应堆压力容器外部冷却（ERVC）是实现熔融物堆内滞留（IVR）的重要方案之一,而反应堆压力容器（RPV）外壁面的临界热流密度（CHF）决定了ERVC冷却能力的限值。为此建立小型CHF试验装置,并采用RPV用SA508钢制作试验块加热表面。以去离子水为试验工质,开展池沸腾下朝向CHF试验,研究真实RPV表面材料在不同倾角和过冷度条件下的CHF特性,及其老化效应对CHF的影响。结果表明：SA508钢表面极易氧化生锈,其CHF较不易生锈的铜和不锈钢表面要高;SA508钢表面CHF随倾角的增大而增加,但在30°附近存在转折,转折角以下范围内的CHF随倾角增加趋势不明显;CHF随过冷度的增加而增加,且基本呈线性变化。本试验有助于进一步认识RPV外壁面的CHF行为,为后续开展CHF增强方法研究奠定基础。     

Modelling of PWR lower head failure under severe accident loading using improved shells of revolution theory

In the study of severe pressurized water reactor accidents, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are usually investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exists. This may lead to an out of vessel steam explosion or to direct heating of the containment; both which have the potential to lead to early containment failure.Within the framework of the OECD Lower Head Failure (OLHF) programme, a simplified model based on the theory of shells of revolution under symmetrical loading was developed by IRSN. After successfully interpreting some other representative experiments on lower head failures, the model was recently integrated into the European integral severe accident computer ASTEC code. The model was also used to obtain the thermo-mechanical behaviour of a 900-MWe pressurized water reactor lower head, subjected to transient heat fluxes under severe accident conditions.The main objective of this paper is to present: (1) the full mathematical formulations used in the development of the model, including their matrices and integrals defined by analytical expressions; (2) the two creep laws implemented, one for the American steel SA533B1 and one for the French steel 16MND5; and (3) the various numerical interpretations of experiments using the simplified model. This paper can be considered as a theoretical manual to aid users of the simplified model during modelling of lower head failures under severe accident conditions. One of the applications presented in this paper concerns the determination of a diagram representing the vessel time to failure as a function of the pressure level and the heat flux intensity. This information has been used by IRSN in probabilistic safety assessment and severe accident management analyses.     

Manufacturing and material properties of ultralarge size forgings for advanced BWRPV

Ultralarge size forgings for the advanced boiling water reactor (ABWR) pressure vessel as represented by the bottom petal made from a 600 ton ingot have been developed. The bottom petal is a larger wall thickness ring with 10 integrated nozzles inside and outside the ring. The outer diameter is 7.8 m, the height is 1.8 m and the wall thickness if 1.1 m in the as-forged condition. A very high purity level of P 0.003% and S 0.003% can be obtained by the application of double-refining processes to all the molten steel. The forging shows a homogeneous chemical distribution, sound internal qualities and adequate impact properties.This paper summarizes the manufacturing technique and material properties of large size forgings such as the bottom petal, the shell with integrated skirt and the bottom dome.     

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。     

Local mechanical properties of Alloy 82/182 dissimilar weld joint between SA508 Gr.1a and F316 SS at RT and 320 °C

The distributions of mechanical and microstructural properties were investigated for the dissimilar metal weld joints between SA508 Gr.1a ferritic steel and F316 austenitic stainless steel with Alloy 82/182 filler metal using small-size tensile specimens. The material properties varied significantly in different zones while those were relatively uniform within each material. In particular, significant gradient of the mechanical properties were observed near the both heat-affected zones (HAZs) of F316 SS and SA508 Gr.1a. Thus, the yield stress (YS) was under-matched with respect to the both HAZs, although, the YS of the weld metal was over-matched with respect to both base metals. The minimum ductility occurred in the HAZ of SA508 Gr.1a at both test temperatures. The plastic instability stress also varied considerably across the weld joints, with minimum values occurring in the SA508 Gr.1a base metal at RT and in the HAZ of F316 SS at 320 °C. The transmission electron micrographs showed that the strengthening in the HAZ of F316 SS was attributed to the strain hardening, induced by a strain mismatch between the weldment and the base metal, which was evidenced by high dislocation density in the HAZ of F316 SS.     

Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

The stress corrosion cracking (SCC) and corrosion fatigue behaviour perpendicular and parallel to the fusion line in the transition region between the Alloy 182 Nickel-base weld metal and the adjacent SA 508 Cl.2 low-alloy reactor pressure vessel (RPV) steel of a simulated dissimilar metal weld joint was investigated under boiling water reactor normal water chemistry conditions. A special emphasis was placed to the question whether a fast growing interdendritic SCC crack in the highly susceptible Alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic SCC crack growth was observed in high-purity or sulphate-containing oxygenated water under constant or periodical partial unloading conditions for those parts of the crack front, which reached the fusion line. In chloride containing water, on the other hand, the interdendritic SCC crack in the Alloy 182 weld metal very easily crossed the fusion line and further propagated with a very high rate as a transgranular crack into the heat-affected zone and base metal of the adjacent low-alloy steel. The observed SCC cracking behaviour at the interface correlates excellently with the field experience of such dissimilar metal weld joints, where SCC cracking was usually confined to the Alloy 182 weld metal.     

J-R fracture properties of SA508-1a ferritic steels and SA312-TP347 austenitic steels for pressurized water reactor’s (PWR) primary coolant piping

J-R fracture resistance of SA508-1a and SA312-TP347 steels, which are both rather peculiar as PWR primary coolant piping materials, were evaluated for application of leak-before-break methodology to the design basis of nuclear power plant piping. Archive materials from various heats of both steel pipes showed apparent heat-to-heat variations in ductile fracture resistance at the operating temperature 316°C. The SA508-1a ferritic steels showed relatively good J-R curve properties although they varied with the microstructures depending on the manufacturing process and chemical compositions. On the other hand, ductile crack growth resistance of SA312-TP347 austenitic stainless steel was unexpectedly poor when carbon content is moderately high. It was found that coarse niobium carbides deteriorated the ductile fracture resistance, so that more rigorous specification is needed in carbon and niobium contents to improve fracture properties of Type 347 stainless steels.     

核压力容器材料国产化的可行性评述

文中扼要介绍了我国首次生产的 RPV 用 A508-3钢锻件的工艺和性能。通过对生产经验、试验研究、国内外文献和国内现有及新添设备的分析给出:实现600MW 核电站压力容器国产化在实际上具备了可行性和现实性。