Features of methods for monitoring the fuel cladding tightness in lead-cooled fast breeder reactors

To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP). The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature. In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.

Application of the Signal Change Tracking Algorithm in the Hydraulic Shock Absorber Monitoring System

During nuclear power plant (NPP) operation, the reactor plant main equipment can show displacements when subjected to the effect of various external and internal loads. These displacements are mainly caused by thermal expansion of the metal and seismic loads. To cope with these phenomena, the reactor plant components that are most susceptible to these types of loads are fastened with hydraulic shock absorbers (HSAs) to limit their displacements under the effect of seismic or accident dynamic loads, as well as to ensure thermal displacements in increasing or decreasing the power unit output. For monitoring the HSA operation and indirectly monitoring the displacements of the reactor plant equipment items fastened with hydraulic shock absorbers, the dedicated hydraulic shock absorber monitoring system (HSAMS) is used, which is equipped with linear displacement sensors installed directly on the HSAs. If the displacements go beyond the predetermined limits, the HSAMs algorithms produce an appropriate alarm. The information from the HSAMS is also used by the automated residual lifetime monitoring system (ARLMS) to calculate the steam generator connection pipe displacement criteria parameters. However, during the operation of a number of NPP power units, a problem associated with numerous failures of the HSAMS linear displacement sensors has been faced. These failures manifested themselves in that the sensor signals went beyond the valid range or frozen under the effect of external influencing factors. As a result, the HSAMS and ARLMS operation was complicated by a large number of unreliable measurements and the functions of these systems were not performed in a proper way. To solve this problem, it has been proposed to use an algorithm for tracking signal changes, which can improve the credibility of HSAMS indications by determining unreliable data in the online mode and by performing statistical processing of the already available array of indications.

Experimental studies into the performance of the lead coolant axial pump wet ends to justify main circulation pumps for the HMLC reactor plant circuits

The paper presents the results of the studies to justify the design solutions for the main circulation pumps of the heavy liquid-metal cooled reactor plant circuits. A substantial difference has been shown in the performance of pumps for the heavy liquid-metal coolant transfer. The studies have confirmed the qualitative difference in the cavitation performance of coolants, the state of the gases and vapors they contain, the influence of supply and discharge devices, and the effects of the impeller blade section performance and geometry and the hub-tip ratio on the pump performance. The studies were performed based on NNSTU’s lead-cooled test facilities with the coolant temperature in a range of 440 to 550 °C and the coolant flow rate of up to 2000 t/h. The outer diameter of the impellers and the straightening devices was about 200 mm, and the thickness of the flat 08Kh18N10T-steel blades was 4.0 mm and that of the airfoil blades was up to 6.0 mm. The pump shaft speed changed in a stepped manner from 600 rpm to 1100 rpm after each 100 rpm. The studies were conducted to justify the engineering and design solutions for pumps as applied to conditions of small and medium plants with fast neutron lead cooled reactors currently under investigation at NNSTU (BRS-GPG). The experimental results can be recommended for use to design other HLMC transfer pumps.

Optimization of the PGV-1000MKO Steam Generator Using CFD

In the VVER -TOI project, new layout solutions were applied in the reactor plant as part of which the steam removal system from the steam generator was changed. Namely, in contrast to the VVER-1000 and VVER-1200 where the steam removal was organized through ten nozzles combined into a steam collector, in the VVER -TOI SG the steam removal was arranged through one nozzle located on the cold collector side. This change leads to the formation of a non-uniform velocity field in the separation volume. To ensure the steam separation characteristics of a horizontal steam generator with one steam nozzle, it was proposed to create a non-uniform resistance on the way of steam motion from the evaporation surface into steam nozzle applying a non-uniform degree of the distribution perforated plate (DPP) perforation. Two computer models of the SG steam volume with different steam removal schemes (one and ten nozzles) were developed, a set of studies on verification and validation was carried out and a set of calculations were performed. Further, to determine the non-uniform degree of DPP perforation, a set of optimization calculations of the SG steam volume with one steam removal nozzle was performed. The non-uniform degree of DPP perforation of the VVER-TOI SG was selected, which provide steam velocity distribution as close as possible to SG with ten steam nozzles. To justify the chosen design, sensitivity analysis was also carried out according to the hole diameters tolerance and steam load profile.

A localization method for loose parts monitoring system of VVER reactor plants

As operational experience shows, it can hardly be excluded that some detached or loosened parts and even foreign objects (hereinafter referred to as the ‘loose parts’) may appear in the main circulation loop of VVER reactor plants. Naturally, the sooner such incidents are detected and evaluated, the more time will be available to eliminate or at least minimize damage to the reactor plant main equipment. The paper describes a method for localizing the impact of loose parts located in the coolant circulation circuit of a VVER reactor plant. To diagnose malfunctions of the reactor plant main equipment, it is necessary to accurately determine the place where the acoustic anomaly occurred. Therefore, if some loose parts make themselves felt, it is important to track the path of their movement along the main circulation circuit as well as their location using physical barriers. The method is based on the representation of the surface, along which an acoustic wave travels, as a 3D model of the reactor plant (RP) main circulation circuit. The model has the form of a graph in which the vertices characterize the control points on the RP surface and the edges are the distances between them. The method uses information about the acoustic wave velocity and the time difference of arrivals (TDOAs) of the signal received by various sensors. It is shown that, when the effect is received by more than three sensors, along with an estimate of the impact coordinate, it becomes possible to estimate the average acoustic wave velocity. To determine time of arrival, the signal dispersion change point detection method is used. Provided that the average size between the control points on the RP surface was 300 mm, the average localization error was about 600 mm. The developed algorithm can be easily adapted to any VVER reactor plant. The obtained deviation values are acceptable for practical use.

Анализ характеристик полифункциональной системы компенсации давления реакторной установки РИТМ-200 атомных ледоколов нового поколения

В настоящее время на ОАО Балтийский завод успешно прошёл ходовые испытания атомный ледокол нового поколения Арктика . Здесь же строятся и находятся на различной фазе готовности два других аналогичных универсальных атомных ледоколов проекта 22220: Сибирь и Урал . На атомные ледоколы данного проекта устанавливаются реакторные установки IY поколения с интегральной компоновкой оборудования, разработанные АО ОКБМ Африкантов и используются другие инновационные разработки, которые ранее не применялись на плавучих объектах. В частности, впервые применяется полифункциональная система компенсации давления в реакторной установке. Практика эксплуатации атомных ледоколов с блочной компоновкой оборудования выявила ряд проблем, возникающих в данной системе. ОКБМ Африкантов приложила значительные усилия для устранения выявленных недостатков. В работе предлагается аналитическое исследование характеристик системы компенсации давления, влияющих на надёжность и безопасность реакторной установки и ледокола в целом. В частности, предложена аналитическая модель для исследования статических характеристик системы компенсации давления при различных режимах её работы. Оценены максимальные относительные отклонения давления в первом контуре при возможных вариантах реализации работы системы, в различных эксплуатационных ситуациях.At present, the nuclear icebreakers of the new generation Arctic has successfully passed the running tests at Baltic Plant. Two other similar universal nuclear icebreakers of project 22220 Siberia and Ural are also being built here and are at a different phase of readiness. Nuclear icebreakers of this project are equipped with IY generation reactor plants with integral layout of equipment developed by OKBM Afrikantov and innovative developments are used, which were not previously used at floating facilities. In particular, for the first time, a polyfunctional pressure compensation system in a reactor plant is used. The practice of operating nuclear icebreakers with a package arrangement of equipment has revealed a number of problems that arise in this system. OKBM Afrikantov has made considerable efforts to minimize these problems. The work proposes an analytical study of the characteristics of the pressure compensation system affecting the reliability and safety of the reactor plant and icebreaker as a whole. In particular, an analytical model is proposed to study static characteristics of the pressure compensation system under different modes of its operation. Maximum relative deviations of pressure in the primary circuit are found at possible variants of system operation provided for by operational situations.