Steam generators – horizontal or vertical (which type should be used in nuclear power plants with VVER?)

The steam generator is a very important component of a nuclear power plant. Historically, vertical steam generators came to be used abroad and horizontal steam generators in our country. Both types of steam generators operate successfully in nuclear power plants and satisfactorily fulfill their functions, enabling the production of electricity. Repeated attempts to re-examine the existing concepts in one or another country have been unsuccessful because there are no convincing arguments for this. Nonetheless, the question of using a different type of steam generator is raised periodically in our country and abroad. This article briefly reviews different concepts of steam generators. Their parameters, characteristics, and thermal efficiency are compared and ways to increase the latter are analyzed. It is shown that it is impossible to choose one or the other type of steam generator without making an exhaustive study and analysis of the layout of the reactor facility and its scheme, servicing, and operation as part of a nuclear power plant. A comparative analysis of layouts of reactor facilities with different types of steam generators is made. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 127–135, September, 2008.     

Upgrading of the sealing units of steam generator in nuclear power plants with VVÉR-440 and-1000 reactors

The results of experimental investigations performed on the sealing units of steam generators in nuclear power plants with VVéR-440 and-1000 reactors during bench tests are presented. The advantages of the upgraded sealing units of steam generator with gaskets made of heat-expanded graphite are presented, and the stages of the development and adoption of the units are indicated. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 470–475, December, 2005.     

Data-driven controller of nuclear steam generators by set membership approximation

Poor control of U-tube steam generators (UTSG) in a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The steam generator is a highly complex, non-linear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. In this paper, a data-driven controller approximated by set membership approach is presented for the water-level control of U-tube steam generators in nuclear power plants. This controller is capable of learning the control action principles from the data obtained using other methods of automatic or manual control. Simulation results of the approximated controller demonstrate its capability in regulating the water level under random disturbances and reference level changes.     

Optimization of the Technical and Economic Performance Indicators of Nuclear Power Plants with Fast Reactors

The results of design analyses for improving nuclear plants with fast reactors, specifically, by using cartridge-vessel generators instead of sectional-modular generators, are presented. It is shown for a nuclear power plant with a BN-800 reactor that the cartridge-vessel steam generators designed by the Special Machine Design Office substantially decrease the metal content, dimensions, mass, amount of construction work, and construction costs of the main vessel of the nuclear power plant.In the BN-800 design, a cartridge-vessel steam generator decreases the specific capital costs for constructing a power-generating unit of a nuclear power plant by approximately 8%, which substantially closes the gap between these costs for nuclear power plnats with BN-800 and VVER-1000 reactors.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 403–412, June 2005.     

Distribution of coolant flow rate in a steam generator in natural coolant circulation regimes

Conclusions The large hydraulic nonuniformity of steam generator pipes operating in parallel in the natural coolant circulation regime results in a lower efficiency of the heat-transfer surface during emergency cooldown of the reactor plant, and it limits the operational possibilities, specifically, for using this regime at partial power levels. It is obvious that circulation reversal in the pipes of steam generators in the natural circulation regime can have an unfavorable influence on individual structural elements of steam generators as a result of additional temperature stresses appearing in the metal. As one can see from Eq. (6), the conditions of the distribution of the coolant flow rate over pipes in a steam generator can be improved at the design stage. Specifically, they can be realized as an efficient ratio of the “macrogeometric” characteristics of the first loop ΔH and H_sgp as well as by the influence on the ratio of the hydraulic resistance of individual sections of the loop, which determine the numerical value of the parameter m. As m increases, other conditions remaining the same, the character of the distribution of the coolant flow rate in the pipes of a horizontal steam generator improves. Thus, designers of a nuclear power plant have ways to search for optimal solutions. It is obvious that the interrelations of the conditions of operation of a steam generator, examined above, and the natural circulation in the loop require that the distribution of the flow rate in a pipe bundle be taken into account in the physical simulation using special thermohydraulic stands. St. Petersburg State Technical University. Translated from Atomnaya énergiya, Vol. 83, No. 3, pp. 169–174, September, 1997.     

Development of a moisture-content system for monitoring leaks in VVÉR pipelines

The main technical decisions adopted in developing a moisture-content system for monitoring leaks in the pipeline of the first loop of VVéR are examined taking account of experience and the main requirements for systems monitoring leaks in pipelines in nuclear power plants. The results of a computational and experimental validation of the serviceability of the diagnostics algorithm and the characteristics of the moisture-content system for monitoring leaks are presented. Using the Kupol-M computer code and a thermophysical stand simulating a leak in a pipeline with basaltic thermal insulation mats in the No. 3 unit of the Kalinin nuclear power plant, it is shown that the system meets the main requirements of the modern concept of leak before rupture. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 291–294, November, 2007.     

Succession in choosing structural solutions for a VVÉR-1500 core

More stringent requirements for VVéR-1500 reactors have made it necessary to introduce innovative solutions into the construction of the core and reactor system. At the same time, succession, i.e., the maximum possible use of tested solutions in developing the VVéR-1500 reactor, decreases the costs and the time required for development, fabrication, and adoption and most importantly it improves reliability and safety. Tested and new solutions for the VVéR-1500 core relative to the VVéR-1500 protoype and possible further upgrading are examined. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 412–416, December, 2005.     

Computational investigation of the effect of hydraulic and nondestructive test parameters on the strength reliability of the heat exchanger tubes of a steam generator

The results of calculations of the probability of a leak appearing in the tube band of steam generator in a VVéR-440 reactor system during operation are presented. The MAVR-1.1 computer code is used to calculate the probability of the formation of a leak and rupture of one of the heat exchanger tubes. The binomial distribution is used to determine the probability of the number of tubes that do not satisfy the plugging criterion. A leak in a tube bank is calculated as a sum of leaks in individual tubes. The probability of such a leak is calculated as a random sum. The calculations show that the parameters of test measures (pressure of the hydraulic tests, reliability of nondestructive testing for defects) and the sequence in which they are performed have a large effect on the failure probability of a tube bank during reactor operation. The computational results and the experience gained in operating steam generators show that the algorithm and the method developed for computing the leak probability could be helpful for estimating the strength reliability of heat exchanger tubes. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 216–221, April, 2007.     

Estimate of the emission rate of radioactive rare gases during an unanticipated accident at a VVÉR-1500 reactor

A method for estimating the emission rate of radioactive rare gases through the exhaust pipes of the passive filtration system in the intershell space of VVéR-1500 and the ventilation pipes of the nuclear power plant is examined. It is proposed that flow and nonflow ionization chambers, which are located either in the exhaust pipes of the passive filtration system in the intershell space of the reactor or in the ventilation pipes of the nuclear power plant, and a spectrometric γ radiation sensor be used to obtain the estimate. __________ Translated from Atomnaya énergiya, Vol. 104, No. 1, pp. 43–54, January, 2008.     

核动力设备耦合优化设计研究

核动力设备重量是评价核动力装置性能的标准之一。蒸汽发生器与稳压器是反应堆一回路中的重要设备,在保证实现其各自功能的前提下,降低这2个设备的重量能提高整个核动力装置的性能。本工作基于秦山核电厂相关设备资料,自主开发了对蒸汽发生器和稳压器进行重量优化设计的计算程序,采用粒子群 模拟退火方法开展多参数优化设计。结果表明,通过参数的重新组合优化,2个设备重量之和减少了18.61%,优化效果显著,相关结果可作为工程设计参考。     

Heavy water reactors and nuclear power plants in the USSR and Russia: Past, present, and future

The history of the development of heavy-water nuclear reactors and the assoiated, installations in the USSR and Russia is presented. Research reactors constructed at the ITEP and under the scientific direction of the ITEP in other countires (Yugoslavia), industrial heavy-water nuclear reactors, and the Maket zero-power reactor are described. Heavy-water gas-cooled reactors for nuclear power plants are discussed in detail: the nuclear power plant with an A-1 reactor, constructed in Czechoslovakia, and the design of maximum-safety nuclear power plant. Electronuclear neutron generators and subcritical nuclear reactors and the possibility of using the for burning weapons plutonium are examined. The electronuclear neutron generator developed at the ITEP is described. State Science Center of the Russian Federation—Institute of Theoetical and Experimental Physics. Translated from Atomanaya énergiya, Vol. 86, No. 4, pp. 310–321, April, 1999.     

Investigation of corrosion degradation of a tube sheet in steam generators in nuclear power plants with vver reactors

Measures for ensuring reliable and safe operation of a tube sheet in the steam generators in nuclear power plants with VVER reactors are examined and validated. The reasons for and mechanisms of the damage occurring to the tube sheets in steam generators are examined. The results of experimental studies evaluating the corrosion effect of the working medium in the second loop on the tube sheet are presented and it is shown that the defects detected by eddy-current monitoring form predominately in the startup and nonstationary operating regimes of steam generators as well as during disruptions of the water-chemistry regime. Suggestions are made for optimizing the water-chemistry regime of the second loop and for preserving a steam generator during downtime. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 195–201, October, 2008.     

CANDU steam generator life management

Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in PWRs. Canadian Deuterium Uranium (CANDU®) steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have resulted in a decrease in steam generator-related station unavailability of Canadian CANDU reactors. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development (R&D) work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for speciality tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service (FFS) guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. This paper will also show how recent advances in cleaning technology are integrated into a life management strategy. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New steam generator designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce-A/B, Pickering-A/B) and strategic plans to ensure that good future operation is ensured. The R&D program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factor.     

Channel reactors as an integral part of nuclear power in Russia

The place and role of channel reactors in nuclear power in our country and the main measures for upgrading and improving the power generating units of nuclear power plants with RBMK reactors are described. It is shown that the risk indicators for serious damage to the core of power generating units with RBMK reactors are lower after upgrading and the corresponding IAEA criterion established for operating nuclear power plants. Upgrading and implementation of a service life extension program has made it possible to obtain licenses for continuing operation of power generating units with first-generation RBMK reactors and predicting a service life increase to 45 years. The characteristics of nuclear power plants with channel reactors with more highly developed internal and natural safety properties are shown in evolutionary designs of the power generating units MKéR-860,-1000, and-1500, which have protective shells and which meet all requirements for power generating units built today. It is shown that innovative solutions for the channel reactor concept can be implemented on the basis of the designs of power generating units with nuclear superheating of steam or on the basis of designs for developing reactors with supercritical parameters. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 29–36, July, 2007.     

The design of the Fort St. Vrain steam generators

The steam generators for the Fort St. Vrain nuclear power plant are the first application in the United States of once-through boiler design coupled with a high temperature gas-cooled reactor. They contain many design features which are unique for this type of component. Since they are an integral part of the primary system and completely enclosed by the prestressed concrete reactor vessel, they must be removable as well as fit the space available for penetrations. These requirements made the once-through boiler principle a logical choice. Multi-start helically wound tubes supported by perforated plates in a star-shape arrangement resulted in an extremely compact design. The helium inlet temperature of 1427°F and steam temperatures of 1005°F main and 1001°F reheat required unique solutions in terms of flexibility and cooling of support systems and selection of insulation materials and design. Operation in a helium atmosphere without a protective oxide layer called for materials with good wear protection characteristics where parts may experience relative motion. Stabilizing orifices, externally adjustable at the steam generator inlet, plus essentially equal tube lengths for each of the many parallel circuits are utilized to balance circuit performance. To minimize gas bypass flows, special gas seals are provided around individual tube bundles. Field erection time was minimized by developing an upper and a lower module assembly and joining them after erection in the reactor vessel.     

Power-generating unit with dual-loop sodium-cooled fast reactor and dual-wall steam generator

The results of a probabilistic analysis performed to validate the safety of AES-2006 designed for the site of the Novovoronezh nuclear power plant are presented. The requirements for the AES-2006 design are examined. The characteristic features of the AES-2006 design for the conditions at the Novovoronezh nuclear power plant site are described, including the diversity of the equipment and operating regime, passive systems, and scheduled maintenance of safety systems with the reactor operating at power. The scope of the probabilistic safety analysis performed at the development stage of the technical design is described. The important problems which must be solved in a probabilistic safety analysis for the designs of new nuclear power plants are discussed. Translated from Atomnaya énergiya,Vol. 106, No. 3, pp. 123–129, March, 2009.     

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.     

Construction,operation, and operating lifetime extension of the steam generators of a BN-600 power-generating unit

The construction of the PGN-200M steam generator of a BN-600 power-generating unit at the Beloyarskaya nuclear power plant is described. Data from 25 years of operation are presented and the basic questions which were solved during the startup-adjustment work, which increased the operational reliability of the steam generator, are elucidated. The advantages of steam generators with different designs are compared, and it is concluded that it is desirable to develop a high-power sodium-cooled vessel steam generator for future facilities. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 481–488, December, 2005.     

Development and investigation of the regulator of the passive heat-removal system for the Kudankulam nuclear power plant (India)

A system for passive removal of the residual heat released in VVéR-1000 with a heat-removal regulator in the form of an air damper operating passively is described. The main results of the computational and experimental validation of the operability of the regulator are presented. It is shown that the design proposed for the regulator gives the nominal characteristics and enables stable operation of the system in hot stand-by and cool-down regimes of the reactor system, and it can be used in the Kudankulam nuclear power plant (India) which is now under construction. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 49–52, January, 2007.