核电站机组故障诊断系统的知识获取

核电站机组故障诊断系统知识的获取和知识库的建立是影响诊断系统能否应用于实际的关键步骤。针对实现核电站机组故障诊断系统给出了知识获取的一种方法和步骤,使其有章可循,加强了在实际核电站中可操作性。按照文中提出的工作框架组织人员完成各项任务,可以最终完成核电站机组故障诊断系统知识库的建立。     

An experimental investigation of a passive cooling unit for nuclear plant containment

A set of condensation experiments in the presence of noncondensables (e.g. air, helium) was conducted to evaluate the heat removal capacity of a passive cooling unit in a post-accident containment. Condensation heat transfer coefficients on a vertically mounted smooth tube have been obtained for total pressure ranging from 2.48×10⁵ Pa(abs) to 4.55×10⁵ Pa(abs) and air mass fraction ranging from 0.30 to 0.65. An empirical correlation for heat transfer coefficient (h), has been developed in terms of a parameter group made up of steam mole fraction (Xs), total pressure (P_t), temperature difference between bulk gas and wall surface (dT). This correlation covers all data points within 20%. All data points are also in good agreement with the prediction of the diffusion layer model (DLM) with suction and are approximately 2.2 times the Uchida heat transfer correlation. Experiments with an axial shroud around the test tube to model the restriction on radial flow experienced within a tube bundle demonstrated a reduction of the heat transfer coefficient by a factor of about 0.6. The effect of helium (simulating hydrogen) on the heat transfer coefficient was investigated for helium mole fraction in noncondensable gases (X_He/X_nc) at 15, 30 and 60%. It was found that the condensation heat transfer coefficients are generally lower when introducing helium into noncondensable gas. The difference is within 20% of air-only cases when X_He/X_nc is less than 30% and total pressure is less than 4.55×10⁵ Pa(abs). A gas stratification phenomenon was clearly observed for helium mole fraction in excess of 60%.     

Quality assurance for objects utilizing nuclear power

The basic aspects of quality assurance for purposes of safety of objects utilizing atomic energy are examined. The formulation of quality assurance requirements of a regulatory agency is analyzed, the experience with enterprises in nuclear power and the atomic industry is examined, and current trends in quality assurance, which are reflected in IAEA manuals and ISO series 9000 standards, are analyzed. The relationship between the quality system and the quality assurance program for objects utilizing atomic energy is discussed. 1 figure, 1 table, 7 references. National Science Center YARB of the Federal Nuclear and Radiation Safety Agency of Russia. Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 230–235, September, 1999.     

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。     

AP1000核电厂反应堆冷却剂系统泄漏分析

第三代AP1000非能动核电厂的主要特征是采用非能动安全原理,使核电厂的系统、设备、构筑物大幅度简化,安全性、可靠性、经济性大幅度提高,以满足美国先进轻水堆业主要求文件的基本要求。本文针对美国业主要求文件(Utility Requirements Document,URD)第三卷第五章《专设安全系统》中对非能动先进轻水堆核电厂反应堆冷却剂系统压力控制功能的要求:在很小的反应堆冷却剂系统(Reactor Coolant System,RCS)净泄漏率(不大于2.27 m3·h-1)条件下,具有足够的系统冷却剂装量及补水能力,以保证在8 h(28 800 s)内不会触发自动降压系统而进行计算分析,本分析采用安全分析报告小破口失水事故(Loss of coolant accident,LOCA)分析采用的NOTRUMP程序,分析结果表明AP1000核电厂可满足上述美国URD要求。     

三门核电AP1000机组辐射防护设计分析

三门核电AP1000机组为第三代核电机组,在辐射防护设计中采用了一回路加锌、较高pH值运行、停堆氧化操作、蒸汽发生器一回路水室电解抛光、优化设备维修、优化屏蔽设计、无线剂量监测等措施,以期降低机组辐射水平和职业照射剂量。本文介绍了三门核电AP1000机组在功率运行及大修期间的辐射水平和职业照射剂量数据,并与国内CPR1000机组的相关数据进行了对比,对AP1000机组的辐射防护设计进行分析,给出了三门核电AP1000机组在辐射防护运行管理及技术改进方面的建议。     

Cladding rupture life control methods for a power-cycling WWER-1000 nuclear unit

Using the cladding creep energy theory, taking into account the WWER-1000 fuel assembly four-year operating period transposition algorithm, as well as considering the disposition of control rods, the location of the axial segment limiting the fuel cladding operation time, at day cycle power maneuvering, has been found. It has been shown that the WWER-1000 fuel element cladding rupture life, at normal variable loading operation conditions, can be controlled by an optimal assignment of the coolant temperature regime and the fuel assembly transposition algorithm.     

Aging management of containment structures in nuclear power plants

Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.     

Quality assurance for a nuclear power plant simulator by applying standards for safety-critical software

Nuclear power plant simulators are playing a more important role in nuclear power plant lifecycle analysis, and the quality of the simulators should be verified to ensure the safety of nuclear power plants. Currently, there is no systematic quality assurance method for nuclear power plant simulators. In this paper, a systematic quality assurance method for nuclear power plant simulators is proposed basing on experiences with safety-critical software. Key aspects of the method are discussed. In addition, application of this method to a real project is also described as a practical reference.     

Failure pressure evaluation of the containment building of a large dry nuclear power plant

The evaluation of the failure pressure of the containment building of a large dry PWR-W three loops nuclear power plant, based on computer numerical simulation, is described in this paper. The proposed method considers fully three-dimensional finite element models in order to take into account the effect of the most significant structural characteristics (presence of three buttresses, penetrations, additional reinforcement around the penetrations, etc.), the lack of symmetry of the forces generated by the prestressing system, as well as the nonlinear behaviour of the materials and the sensitivity of the results to uncertainties associated with several parameters. The computational model is completely described, including the constitutive equations for the concrete, the reinforcing steel and prestressing tendons, the spatial discretization—isoparametric elements including the reinforcement are used. The structural models and the analyses performed for their calibration are also described. The influence on the failure pressure of incorporating the foundation slab in the structural model, and the influence of the thermal effects, are discussed. One of the conclusions of the numerical study is that the failure process can be appropriately simulated by means of a structural model which does not include either the foundation slab or the thermal effects. Finally, results of a probabilistic simulation of the failure pressure are given.     

Method for assessing the consequences of radiation accidents in the rooms of a power-generating unit of a nuclear power plant with a VVÉR-1000 reactor

Monte-Carlo estimation of the dose rate of external radiation from volume and surface sources of a gas-aerosol radioactive impurity, arising in the rooms of a power-generating unit during a radiation accident at a nuclear power plant, is examined. The volume and surface sources are distributed uniformly. The radial distribution of the dose buildup factor and the dose rate are obtained as functions of the γ-ray energy at different heights. The calculations are performed for distances no more than four γ-ray mean-free path lengths (μd ≤ 4) by the local flux estimation method. Comparisons with similar calculations performed with computer programs employing the discrete-ordinate method show satisfactory agreement. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 254–262, April, 2007.     

Thermohydrodynamic calculation of the subdome part of the containment envelope of a nuclear power plant

Main Scientific Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 78, No. 2, pp. 172–176, March, 1995.     

Choice of the unit power-generating capacity of a nuclear power plant

A characteristic of the present status of nuclear power in Russia is that in the next few years it will be necessary to make basic technological and economic decisions that will have long-term consequences. These decisions must concern all aspects of the nuclear-power complex. Specifically, at the present time there is no validation of the present and future requirements for the capacity of serially manufactured power-generating units of nuclear power plants with VVER or fast reactors. The problem of choosing the unit capacity of a nuclear power plant must be examined taking account of different factors and not solely from the standpoint of minimizing the capital and operational components of the cost of electricity. The main objective of this work was to develop recommendations for validating the optimal capacity of powergenerating units in nuclear power plants (capital costs, construction time, harm due to unanticipated stoppage of the power-generating units, unification, manufacturing quality, harm due to accidents, and so forth), the possibilities of electric grids, and the regional demand for electricity. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 243–248, November, 2008.