Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models.This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 Nuclear Power Plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. Some of the proposed modifications of the system, which would significantly decrease risks associated with accidents that are beyond the original design basis, are discussed. The design of the newer 213 model differs in many aspects. It incorporates the simple and original application of passive natural processes to limit the large-break loss-of-coolant accident post accident pressure. Other features of this containment design, such as complicated geometry, dependence on several mechanical devices and interlocks, and insufficient experimental evidence, lead to doubts concerning the operation of this containment under accident conditions. For the newer 213 model, current work is devoted mainly to safety assessment and verification of the containment design. Some information concerning the on-going work is provided in the paper.     

Modeling of atmosphere stratification in containments of nuclear power plants using lumped-parameter code

One of the current high priority safety issues of nuclear power plants is atmosphere stratification in the containments. It requires extensive experimental and numerical investigations. Numerical investigation of stratification using lumped-parameter codes is complicated due to inherent limitations of the codes. These limitations have to be taken into account in developing nodalisation of the containments.The paper presents M5 experiment simulations using lumped parameter code COCOSYS. Experiment M5 was performed in the MISTRA test facility in the frames of FP-6 project SARNET (Severe Accident Research Network). Obtained results are presented in the context of evaluation of the lumped-parameter code COCOSYS capabilities to model the formation of stratified atmosphere and to give recommendations for the development of nodalisations in similar cases using any lumped-parameter code. The paper describes the MISTRA test facility and the performed experiment M5, the gas flow model of the COCOSYS code, the developed nodalisations, and the obtained results. The performed analysis showed that detailed vertical nodalisation and special treatment of injected gas flow is required for simulation of gas mixing phenomena.     

Inerting and venting of mark II containments in BWR nuclear plants

An analysis of hydrogen control systems corroborates containment inerting as the only way of preventing hydrogen explosions which may jeopardize the integrity of BWR Mark II containments during severe accidents. A severe Large Break LOCA and a severe Stuck Open Relief Valve Accident are simulated by the MARCH 2.0 code to compare the advantages and disadvantages of pre-inerting and post-inerting, with or without venting, in BWR Mark II containments.     

Mechanical and leakrate predictions for nuclear power plant containments in accidental conditions

In 1996, EDF decided to build a containment model at the scale 1:3, the Maquette echange vapeur/air (MAEVA) mock-up, in order to check and study the behavior of a prestressed concrete containment vessel without liner in terms of mechanical strength and leaktightness, for loadings corresponding to its design and beyond design conditions. In parallel to the construction and testing of the mock-up, predictive calculations of the mechanical and leaktightness behaviour of the mock-up were performed in the framework of a cost shared R&D action supported by the European Union, the Containment Evaluation under Severe Accidents (CESA) project. The strategy of EDF concerning the R&D performed on leaktightness of concrete and concrete structures is first explained and the goal of two interesting programs is shortly presented and discussed in the first part of the paper. Then, the emphasis is made on the predictive calculations performed on the MAEVA mock-up by means of finite elements (FE) calculations. A summary of the main achievements is then given and the interest of FE calculations is discussed for describing both the mechanical and leaktightness behaviour of a concrete structure as a function of crack development.     

Welding consumables for nuclear power plants

In nuclear power plants, submerged arc welding and covered arc welding have long been employed especially for main weld seams, including the core region of RPV.This paper investigates the mechanical properties of several welding consumables we have developed for industrial plants — that is, welding consumables which lower the phosphorus and copper content of the welded metal, those for plates possessing particularly high tensile strength and those for the narrow gap welding method.Recent data derived from irradiation embrittlement tests show that these welded metals using a non-copper coating are highly effective in minimizing shifts in the transition curve.Welding consumables for A533B C1.2, A543 C1.1 or A508 C1.4 steels have a higher tensile strength than those for A533B C1.1 or A508 C1.3.We have developed submerged arc and covered arc welding consumables to be used with these kinds of steels, and it was confirmed that these consumables possess excellent tensile strength and notch toughness.Our tests also confirmed that the narrow gap SAW and MIG welds are more efficient than the conventional ones. Moreover, the mechanical properties of the welded metals are also excellent.     

Flood hazards for nuclear power plants

Flooding hazards for nuclear power plants may be caused by various external geophysical events. In this paper the hydrologic hazards from flash floods, river floods and heavy rain at the plant site are considered. Depending on the mode of analysis, two types of hazard evaluation are identified: (a) design hazard which is the probability of flooding over an expected service period, and (b) operational hazard which deals with real-time forecasting of the probability of flooding of an incoming event. Hazard evaluation techniques using flood frequency analysis can only be used for type (a) design hazard. Evaluation techniques using rainfall-runoff simulation or multi-station correlation can be used for both types of hazard prediction.     

Securing safe operation of VVÉR reactor systems in nuclear power plants

The safe operation of VVéR reactors has been discussed throughout the entire design process, taking account of the normative documentation, including the international requirements (IAEA, EUR). After the first domestic normative document “Basic principles for securing the safety of nuclear power plants” was approved in 1973, work began on the reconstruction of the first-generation VVéR-440 power-generating units. The measures taken to increase safety concerned all types of reactors VVéR-440 and-1000. Information on implementing these measures is presented. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 87–93, August, 2006.     

Offshore nuclear power plants

Offshore Power Systems, a joint enterprise of Westinghouse and Tenneco, has been formed to manufacture floating nuclear power plants. Commitments for the first two offshore plants have been received from the Public Service Electric and Gas Company. This paper describes the floating nuclear plant concept with special reference to its advantages and its novel features. The novel features are a consequence of the floating aspect and include the design of the platform, the safety analysis and also the analysis and specification of plant motions due to environmental effects such as wind, waves and earthquakes. Site-related aspects such as the breakwater and mooring systems are discussed. The nuclear power plants will be manufactured in a central facility and this manufacturing concept is described.     

Alexisismon isolation engineering for nuclear power plants

This paper reviews the special requirements regarding efficiency, licensibility (reliability) and cost which should be met to achieve an optimum base isolated nuclear power plant design. It then describes the Alexisismon-2, patented isolation system developed by the author, underlines its original properties (linearity and separation of functions) and presents a conceptual design of its application to a nuclear power plant. The great reliability of the system components is demonstrated. The efficiency of the A-2 is found to be very high: a reduction factor of the base shear induced in the plant higher than 25 is achieved for all examined real accelerograms scaled to 1 g GPA. So the isolation components, the structural system of the plant, its equipment and systems can be easily designed to remain in the elastic range of stresses and strain even for seismic input with GPA higher than 2 g.     

Unitized concept for earthquake-resistant nuclear power plants

This paper discusses a unitized concept for an earthquake-resistant nuclear power plant which can withstand major earthquake shaking and fault slips without releasing radioactive material into the atmosphere. A 1000 MWe pressurized water reactor power plant of recent design is adapted to a unitized concept, and cost studies are made for the incremental cost.     

Work on nuclear power plants for space

Conclusions It should be noted that the space problems gave an unprecedented impetus to improving computational methods and design of reactors and shielding and the development of the latest technology. This is not surprising, since the designs required that the mass of the nuclear power plant be determined to within several kilograms. The placement of every kilogram of a satellite in orbit costs several thousands of dollars. The development of nuclear rocket motors and the first thermionic systems "Topaz" in the world was a very important achievement in space nuclear technology. Unfortunately, as sometimes happens in practice, this achievement was far ahead of its time and remains unused. Translated from Atomnaya énergiya, Vol. 80, No. 5, pp. 357–361, May, 1996.     

核电站环境放射性监测

随着国家核电产业的发展,越来越多的核电厂将陆续建设,为了更好的保护环境、保护公众,判断核电厂对环境、对周围居民的影响,环境放射性监测是非常必要的。文章就核电厂环境放射性监测的目的,监测项目,监测方法等方面进行介绍。     

Cosmic thermoemissional nuclear power plants

Translated from Atomnaya Énergiya, Vol. 69, No. 4, pp. 215–219, October, 1990.     

Early-age behaviour of concrete nuclear containments

A numerical model has been developed to predict early-age cracking for massive concrete structures. Taking into account creep at early-age is essential if one wants to predict quantitatively the induced stresses if autogenous or thermal strains are restrained. Because creep strains may relax internal stresses, a creep model which includes the effects of hydration and temperature is used. For the prediction of cracking, a simple elastic damage model is used. Numerical simulations are performed in order to predict the behaviour of a massive wall and a concrete containment of a nuclear power plant. They show that significant relaxation of stresses (due to creep) occurs only after about 10 days, after cracking occurs. Moreover, since temperature in concrete may reach important values in massive concrete structures, it appears that effect of temperature on creep must be taken into account for an accurate prediction of cracking.     

Dynamic analysis for design criteria for underground nuclear reactor containments

This paper is based on i) the recent input of the authors for the Underground Containment Sub-section of the Seismic Task Group Report of the ASCE Committee for Nuclear Structures and Materials, and ii) parametric studies carried out by the first author on the principal underground concepts.The extensive work on aseismic design of above-ground reactors and recent studies on missile impact effects, aircraft impact, blast effects due to chemical explosions, reactor core melt-down and tornadoes indicate the advantages of underground siting with inherent general reduction to complexity of seismic amplification and benefits of structural and biological integrity. Other advantages are possibilities of urban siting, ecological considerations, reduced effects on the landscape, ability to design three-dimensionally, separation of component facilities, support capability to equipment, reduced power transmission costs, increased number of acceptable units and power capability from a single location, and reduction of decommissioning problems.In view of the limited actual experience in the structural design of underground containments (only four European reactors), the proposals are based on a) the transposition of applicable design specifications, constraints and criteria from existing surface nuclear power plants to underground, and b) the use of many years of experience in the structural design of large underground cavities and cavity complexes for other purposes such as mining, hydropower stations etc. All concept options are assumed to be similar in design criteria for structural competence to contain radioactivity and fuel heat and meet the functional, servicing, protective and aesthetic requirements. The choice of underground siting should be based on criteria developed from the sequential consideration of load-causing phenomena, concept and site characteristics. From the criteria, loads for a particular concept and site are then calculated and the design formulated.The state-of-the-art is presented and guidelines are outlined for 1) Load causing phenomena, 2) Underground siting concept considerations, 3) Siting factors and structural selection, 4) Structural types, 5) Analysis (including comparison of lumped parameter and finite element analysis), 6) Design procedures, and 7) Miscellaneous considerations (laboratory tests and field measurements, secondary equipment, faults, leakage of radioactivity, ground water control, environmental factors).Parametric studies are described for structural characteristics of the four principal underground concepts: a) Cut-and-cover in rock or soil, b) Unlined cavity in rock, c) Lined cavity in rock or soil, and d) Lined cavity in rock or soil with annular filling of soft material — with respect to shape, backfill material, cavity wall reinforcement, passive and active rock bolting, lining and annular filling. The response to a step pulse, representing a blast excitation applied horizontally, is studied. As the character, intensity, duration and frequency of earthquake and blast-induced ground motions are roughly similar, the results have practical value in studying earthquake effects.