M2N nitride phases of 9% chromium steels for nuclear applications

M₂N nitride phases of 9% chromium steels with an extra-low carbon content have been investigated using a transmission electron microscope and an energy-dispersive X-ray (EDX) spectroscopy. The steel samples were normalized for 1 h at 1050 °C and then tempered at 600-780 °C for 30 min to 5 h followed by an air cooling. Through the analyses of the electron micro-diffraction patterns and EDX data for the precipitate particles on the extracted carbon replica, two types of Cr-rich M₂N nitride phases with the same hexagonal structure but totally different lattice parameters, a = 2.80 Å/c = 4.45 Å and a = 7.76 Å/c = 4.438 Å, were determined in the steels. Four types of Cr-rich M₂N phases with different lattice parameters probably existed in the steels. The M₂N phase revealed a decrease in its Cr content, an increase in its V content as the tempering temperature was increased, and no obvious change in its content for the metal fraction with an increasing tempering time.     

TEM investigations of MN nitride phases in a 9% chromium ferritic/martensitic steel with normalization conditions for nuclear reactors

The investigations on the precipitate phases in a 9%Cr ferritic/martensitic steel under different normalization conditions have been made by using a transmission electron microscope and an energy-dispersive X-ray spectroscopy. Hot-rolled steel samples were normalized at 1050-1200 °C for 1-2 h followed by an air cooling to room temperature. MN vanadium nitride precipitates with a plate-like morphology and a chemical formula of about (V_0.4Nb_0.4Cr_0.2)N have been observed at triple junctions, grain boundaries and within matrix in the steel samples normalized at 1050-1150 °C for 1-2 h, but they were dissolved out at 1200 °C within 1 h. Vanadium nitride is a stable phase at 1050 °C according to thermocalc prediction of equilibrium phases in the steel. With increasing normalizing temperature and time, there was no a striking change in the chemical composition of metallic elements in the MN phase, but a considerable increase in the size of the MN precipitate.     

Precipitate phases of a ferritic/martensitic 9% Cr steel for nuclear power reactors

One of the reasons that ferritic/martenstic steels have been considered as candidate materials for nuclear power reactors is their superior creep resistance at elevated temperature. The creep rupture strength of 9% chromium steel could be improved by a fine dispersion of secondary precipitate phase. The precipitate phases in extra-low carbon 9% chromium steel with tempered conditions were investigated by transmission electron microscope and energy-dispersive X-ray analysis. The steel specimens were normalized and then tempered at different temperatures. Niobium-rich MN nitrides (Nb_0.6V_0.3Cr_0.1)N, and two kinds of vanadium nitrides, (V_0.6Nb_0.2Cr_0.2)N and (V_0.45Nb_0.45Cr_0.1)N having a f.c.c. crystal structure, were identified in the steel specimens tempered at 600-780 °C, and 750 or 780 °C respectively. Hexagonal chromium-rich M₂N precipitate phases with different lattice parameters, a = 2.80 Å/c = 4.45 Å and a = 7.76 Å/c = 4.438 Å, were determined in the tempered steel specimens. The M₂N phase showed a decrease/an increase in its chromium/vanadium content as the tempering temperature was increased. The influence of precipitates and heat treatment conditions on the high temperature properties of 9% Cr steel was discussed.     

Recent applications of PSA for managing nuclear power plant safety

The safety design and regulation of nuclear power plants has traditionally been based upon deterministic approaches that consider a set of challenges to safety, e.g. design basis accidents, and determine how those challenges should be handled. The approach has been very successful since no plant designed or regulated to United States standards has ever harmed a member of the public. The arbitrary nature of these safety criteria, the potential inconsistencies in the judgments on relative probabilities, and the lack of definition for ‘safety’ became increasingly evident during the 1960s. Probabilistic approaches to reactor safety were proposed ^1,2,3 but did not take off in the United States until publication of the Reactor Safety Study ⁴ in 1975. Even as the methodology matured, there remained a challenge to integrate it into the regulatory process. This article will describe this integration process. A probabilistic approach to regulation enhances and extends the traditional deterministic approach by introducing the concept of safety (risk) significance that allows the designer/operator to focus on important issues. Emphasis was initially placed on relative risk but now regulatory decision-making is employing both relative and absolute risk. Measures of importance will be defined. Risk information can be used to prioritize the allocation of resources and three examples will be described. Equipment configuration control systems are being installed and used at nuclear power plants to enhance safety and to reduce Operating and Maintenance costs; they will be described. Finally, the US Nuclear Regulatory Commission's introduction of risk-informed decision-making into the regulatory process will be discussed.     

The new Cyclone 18/9 beam transport line at the CNA (Sevilla) for high energy PIXE applications

The Cyclone 18/9 cyclotron system at the Centro Nacional de Aceleradores (Sevilla, Spain) is commonly used to create short life radioisotopes for PET applications. Besides, an external beam transport line has been recently installed in one of the target ports with two major purposes: to study the effects of 18 MeV protons irradiation on the behaviour of electronic devices for space applications and to complement the analysis of materials using our 3 MV tandem accelerator with the PIXE measurements at high energy.In this work, the main elements of our beamline will be briefly described and the first PIXE application will be presented. The usual PIXE, in the analysis of archaeological metallic objects, using around 3 MeV protons requires having a shiny area. Our purpose is to obtain a deeper determination of the bulk composition bombarding with 18 MeV protons through the corroded samples surfaces, without polishing the ancient object. To check this methodology high energy PIXE has been performed on two fibulae of the Later Bronze Age and First Iron Age, coming from the area around Sevilla.     

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.     

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.     

核电厂HRA定性分析

人因可靠性分析(HRA)是核电厂概率安全评价(PSA)的重要组成部分,定性分析是HRA的基础和出发点。本文介绍了核电厂HRA定性分析的目的、原则、方法和程序,并以压水堆核电厂SGTR为具体实例进行说明。     

Principles for arranging dose-rate meters around a nuclear power plant

All-Union Scientific-Research Institute of Nuclear Power Plants. Translated from Atomnaya énergiya, Vol. 76, No. 3, pp. 188–193, March, 1994.     

Structural response of a nuclear power plant steel containment under H2 detonation

At present studies are in progress at ENEA-DISP to assess the performance of a steel containment under hydrogen detonation.In this framework a number of topics needs to be investigated and they are considered in the present paper. To get a better understanding of the containment wall behaviour under a detonation a simple but complete model is analysed in order to study the fluid-structure interaction during the explosion. The structure is represented by a single degree of freedom (SDOF) elastic-plastic system. This system is coupled to a monodimensional model of the containment atmosphere excited by hydrogen bursting. The atmosphere modeling allows to represent the shock propagation and the reflected wave effects. In the model a cylindrical geometry is used as reference. The obtained results are compared with data adopted in Italy to assess the structural integrity of the Alto Lazio NPP steel containment in the case of a severe accident.The limits of the model as well as the possible extensions are discussed in the paper together with a possible application in an experimental program directed to the assessment of failure criteria under severe accident conditions.     

The early stages of the oxidation of a 9% chromium steel em carbon dioxide studied by Auger electron spectroscopy

The early stages of the oxidation of 9% Cr steel in carbon dioxide at atmospheric pressure and 773 K have been studied by use of Auger electron spectroscopy. The surface of the sample was monitored throughout the oxidation process without changing the environment other than to evacuate the system for a short period so that the Auger spectrum could be recorded. Sulphur was found to be an important participant in the early stages of oxidation and an oxide film containing iron was formed. Chromium was not detected in this oxide, indicating that its presence in the alloy is not conducive to the formation of a protective Cr₂O₃ layer. The mechanism of oxidation was similar to that of pure iron under the same conditions.     

Emergency management for a nuclear power plant using fuzzy cognitive maps

This paper explores the application of fuzzy cognitive maps (FCM) to emergency operating procedures (EOPS), to represent the decision-making process during abnormal situations in a nuclear power plant (NPP). The decision-making process in a NPP is a complex process, due to the many elements involved in its operation, and the permanent attention demanded by its maintenance. At the present time, the decision making process in a NPP is analyzed and developed by reactor operators, based on a set of instructions as well as flow charts to mitigate the consequences of a broad range of transients, accidents and multiple equipment failures, whose main characteristic is to be linear representations of events within a scenario. One of the main objectives of this paper is to present a method based in FCM that could be applied in the development of EOPS, and show some simulations, specifically the loss of coolant accident (LOCA) scenario in a boiling water reactor (BWR) with the Mark II containment design was studied. The FCM-based method represents with high fidelity the expert reasoning (the human expert is very important) and the interpretation of the results aids instantly to the reactor operators in the surveillance of the reactor proper functionality due that they have the responsibility of the decision taking in emergency situations. The simulations results show that the FCM predict properly the phenomenon in the reactor vessel and primary containment.     

秦山第二核电厂阀门类设备弱贯穿辐射调查

在秦山第二核电厂8次大修期间,对反应堆冷却剂系统(RCP)、余热排出系统(RRA)、化学和容积控制系统(RCV)、反应堆换料水池和乏燃料水池冷却和处理系统(PTR)4个系统主要阀门的辐射源项和弱贯穿辐射进行了监测。测得RCV系统阀门沉积的放射性核素主要是^110mAg,RCP、RRA和PTR系统阀门内沉积的主要是⁵⁸Co、⁶⁰Co、⁵¹Cr、⁹⁵Nb、⁹⁵Zr等放射性核素,伴随的β射线能量主要在500 keV范围内。测量给出了4类阀门的公式′(0.07)和H·′(3)值,测得H·′(3) /H·^*(10)值在1.24左右,H·′(0.07) /H·^*(10)值在14左右。结合测量结果,给出了部分阀门需要对检修人员开展眼晶体剂量和皮肤剂量监测的建议。     

1000MW核电管板纯净钢锻件制造工艺及其性能

根据1000MW核电管板用纯净钢锻件性能和组织的要求,利用合金化原理确定了冶炼时钢中各合金元素的成分控制方向;采用电炉加钢包炉加真空浇注进行冶炼浇注,真空浇注过程中加保护防止二次氧化;采用特殊的镦粗工艺避免了管板镦粗过程心部产生超标缺陷;采用合理的热处理工艺,保证管板锻件的组织和性能。经检验表明,锻件用钢的质量达到了纯净钢的要求,管板锻件的综合性能达到世界领先水平。     

某核电厂核事故应急演习经验反馈

日本福岛核事故的发生对全球范围核电行业的核应急演习提出了更高要求。本文结合电厂工作实际,就福岛事故后核应急演习新要求及经验反馈进行了总结,供从业人员参考。     

Equipment for vitrification of nuclear power plant wastes in a direct-heating ceramic furnace

Joint-Stock Company ‘Sverdlovsk Scientific-Research Institute of Chemical Machinery.’ N. A. Rakov Ministry of Atomic Energy of the Russian Federation. Translated from Atomnaya énergiya, Vol. 80, No. 3, pp. 219–221, March, 1996.     

Application of concentrated ozone for decontamination of equipment in a nuclear power plant

The destruction of chromium oxides on the surface of 12Kh18N10T steel by concentrated ozone at different temperatures is investigated. The optimal temperature of the process in water is found to be 95°C. The oxidation time of the surface layer of chromium oxides is 5–10 min. Subsequent ozone treatment of a steel surface for a 20-fold longer time did not show any substantive destruction of the surface of corrosionresistant steel. Ozone diffusion through the surface of a bubble to the metal surface plays a decisive role in the oxidation of chromium.     

Flux measurements in a nuclear research reactor by using an aluminum nitride detector

A small polycrystalline aluminium nitride detector with a thickness of 381 μm was used to measure a 200,000 Ci Co⁶⁰ source and to measure the flux in a research reactor where the neutron flux is about 10¹⁴/cm² s, which is nearly the same order as in the commercial power plant. If the applied voltage is greater than or equal to 2000 V and if the measurements are done in a short period of time so that the heat energy does not build up in the aluminium nitride, then the measured electric current is linearly proportional to the input flux. It is assumed of course that the energy spectrum of the input flux remains constant. This linearity relation is illustrated by the results of a measurement in which the reactor power has been controlled so that the flux becomes a step function.