VVER-1000 Reactor Research Articles

ПРОВЕДЕНИЕ СРАВНИТЕЛЬНЫХ РАСЧЕТОВ ИЗМЕНЕНИЯ НУКЛИДНОГО СОСТАВА ЯДЕРНОГО ТОПЛИВА ПРИ ВЫГОРАНИИ

Features of calculation methods implemented in computer codes, evaluated nuclear data libraries and computer models have a significant impact on the results of burnup calculations. This article contains comparative calculations of VVER-1000 reactor nuclear fuel burnup and demonstrates an influence of different factors on the results of these calculations. The results of comparative analisys demonstrate the necessity to intensify creation of benchmarks with experimental data on nuclide concentarions in fuel including fuel for fast neutron reactors.

Radiation tests of products made of calcium-thermal zirconium grade СTZ-110 under operation of the VVER-440 reactor

The paper presents the results of reactor tests of fuel assemblies with cladding made of zirconium alloy grade СTZ-110 at the material testing reactor of National Research Center (NRC) "Kurchatov Institute" at Novovoronezh and Leningrad NPP under various nuclear fuel burnups. It was shown that after all test cycles, the parts from zirconium in the fuel assemblies were in good condition, which was confirmed by metal research of the samples cut out from the fuel elements. The mechanical properties of the fuel cladding made of СTZ-110 alloy are stable and satisfactory. Maximum burn-outs were achieved in the VVER-1000 mode of 67.4 MW·day/kg of uranium and the RBMK-1000 mode – 76.0 MW·day/kg of uranium.

Development of canisters for spent fuels of VVER-1200 and ATMEA1 new generation reactors and determination of geological disposal densities

Geological disposal is a method widely accepted as safe and feasible for the permanent disposal of spent fuel. There is not yet an operating geological disposal facility, however, there are various projects ongoing in many countries. KBS-3 disposal concept developed by Sweden is the basis of most of the geological repositories under development but there are differences among the concepts mainly in the geometry and material of disposal canisters, geological formations of host rock and vertical or horizontal emplacement of canisters. The aim of this study is to design disposal canisters for spent fuels of two representative new generation reactors, VVER-1200 and ATMEA1, and determine the disposal densities of designed canisters when emplaced horizontally in a granitic rock formation. In the first part of the study, characteristics of spent fuels discharged from VVER-1200 and ATMEA1 reactors and cooled for various time periods before disposal are determined by utilizing the MONTEBURNS 2.0 code. Then, canister designs for different spent fuel loading patterns (various spent fuel numbers, geometric orientations, pre-cooling periods) are developed by taking into consideration the dimensional and thermal limitations. In the last part of the study, suitable disposal distances between the designed canisters are determined for horizontal emplacement in a geological repository by performing thermal analysis in the ANSYS and most favorable canister designs for VVER-1200 and ATMEA1 spent fuels are determined with respect to disposal density. Results showed that canisters consisting of 5 VVER-1200 spent fuel assemblies with 60 years pre-cooling period and 5 ATMEA1 assemblies with 70 years pre-cooling period require less spacing and disposal area than other canisters investigated.

Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2019

Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2019

Stresses Induced by LOCA in Steam Generator Collector of VVER1000

During Double End Break Loss of Coolant Accident (DEBLOCA) in VVER1000 reactor, in the first second, pressure waves had been noticed in all points hitting the main components of the primary circuit. Studying the stresses due to DEBLOCA is important to avoid equipment failure. The main attention is focused on the pressure drops on the equipment, because this parameter determines the dynamic loads on the equipment, which can lead to its failure. ATHLET code is used to obtain pressure and temperature variation with time during the accident for the VVER1000 reactor Kalinin-3. The results are used in ANSYS Finite element package to calculate stresses in steam generator collector. The stress just before the emergency (at time 0.014 sec) was 4.3e7 pa and becomes 4.38e6 pa just after the emergency (at time =0.022 second). Fluctuations appear with two peaks of highest amplitude of 1.87e7 and 2.38e7 pa at time of 0.06 and 0.094 sec respectively which is well below the allowable stress. The results indicate safe operation of the equipment.

Digital technologies in nuclear power plant probabilistic safety assessment

The ability of systems, structures and components that provide ultimate heat sink, to perform their functions and to ensure the safety of the nuclear power plant with VVER-1200 reactor in the events of external influences that entail the loss of ultimate heat sink is subject of the study. Necessity of consideration of mentioned events was demonstrated by last events occurred on NPPs worldwide in particular in Fukushima-Daiichi accident. Probabilistic model was developed implementing modern methods for assessment of probabilistic safety values using digital software RiskSpectrumPSA. Calculated probabilistic safety values confirm chosen design solutions and possibility of probabilistic safety assessment application for similar tasks.

The Impact of Cesium Vapor Bulk Condensation on the Transport of Aerosols in the Reactor Plant’s Primary Circuit during a Severe Accident at a Nuclear Power Plant with VVER

The article presents the results from numerical modeling of aerosol particles' deposition process during their transportation in the horizontal section of the reactor plant’s primary circuit model for the conditions typical for a severe accident at an NPP equipped with VVER reactors. The effect that the bulk condensation of cesium vapors has on the growth and deposition of aerosols during their transportation is analyzed. The influence of the initial size and concentration of aerosol particles on their growth rate and subsequent deposition is considered. Homogeneous condensation occurring in the absence of heterogeneous condensation nuclei (solid aerosol particles) and also heterogeneous condensation under the conditions in which nuclei of the new phase do not emerge according to the homogeneous mechanism are considered as the limit cases. Data on the average size of the produced droplets and aerosol particles along the channel axis in the condensation zone with the varied value of aerosol particles mass fraction in the flow have been obtained, which corresponded to different ratios of the condensate mass fraction in the form of droplets and liquid layer on the surface of solid aerosol particles. It has been established that the general deposition rate decreases if the flow contains droplets generated as a result of homogeneous bulk condensation and having a size smaller than the size of grown aerosol particles. The article describes the analysis scheme of the modeled section whose geometrical shape was specified based on the horizontal section of the experimental installation in the FALCON-ISP1 series of experiments on the deposition and transport of radionuclide aerosols. The cesium vapor homogeneous-heterogeneous condensation process is modeled by numerically solving—using the method of moments—the kinetic equation for the droplet size distribution function with involvement of the authors’ COND-KINET-1 computer program. The transport and deposition of aerosol particles in the primary circuit channel were modeled using the MAVR-TA computer code developed at the National Research Center Kurchatov Institute.

Разработка критерия для регистрации фактов выноса топлива из негерметичных твэлов во время работы реакторов ВВЭР

Разработка критерия для регистрации фактов выноса топлива из негерметичных твэлов во время работы реакторов ВВЭР

Использование эрбия в качестве выгорающего поглотителя в реакторах типа ВВЭР при работе на удлиненных кампаниях

Использование эрбия в качестве выгорающего поглотителя в реакторах типа ВВЭР при работе на удлиненных кампаниях

Investigation of the VVER-1000 reactor pressure vessel neutron fluence and displacement per atom using MCNP6

Ensuring the integrity of the reactor pressure vessel (RPV) during the operating life of nuclear reactors is of paramount importance, and hence the harmful effects of irradiation on the steels of the RPVs have been extensively studied especially the irradiation embrittlement effects. However, improvements in the calculation methods are still necessary to further reduce the discrepancy between calculations and experiments. This study was performed to examine the neutron fluence and respective rate of displacement per atom (DPA), an important parameter describing the irradiation-induced damage, in the VVER-1000 RPV steel based on the criticality calculation, instead of the common fixed source approach, with the MCNP6 Monte Carlo code, and identify the most sensitive locations in the VVER-1000 RPV. The MCNP6 modeling of the VVER-1000 reactor was verified against the OpenMC calculations in addition to the confirmation of the convergence of the effective neutron multiplication factor and the Shannon entropy of the fission source distribution in the MCNP6 criticality calculations to ensure reliable tally results of the neutron fluence. Consequently, the maximum neutron fluence in the VVER-1000 RPV was found at the positions adjacent to the peripheral fuel assemblies, and the respective DPA rate was analyzed in correlation with the neutron fluence and energy spectrum.

Time-invariant reliability assessment for a passive heat removal system based on FORM methodology with a new approach for finding the design point

Passive systems are widely becoming used in the nuclear industry to improve the safety level of nuclear power plants. However, the functioning parameters of the passive systems are extremely tied to uncertainty, and the reliability assessment of these systems is one of the important open issues in the passive safety system analysis area. So far, there is no common method to estimate the reliability of passive safety systems. The FORM method with the HL-RF algorithm is a popular iterative method to gain reliability index. However, when the nonlinearity of a system increases, the method may cause numerical instability. In this study, the reliability methodology for passive safety systems (RMPS) is applied to evaluate the reliability of the passive heat removal system (PHRS) of the VVER-1000 reactor. A simple and efficient reliability algorithm is proposed to overcome the numerical instabilities of the HL-RF algorithm of FORM based on the effective search direction vectors with an appropriate step size to find the design point. The accuracy of the proposed algorithm is shown through the stochastic response surface method (SRSM). Results show that the proposed algorithm is simple, reliable, fast, and can find the reliability index with high efficiency and accuracy.

Analysis of the Kalinin-3 coolant transient benchmark by SKETCH-N/SKAZKA code system

The switching off of the one of four main circulation pumps results in the Kalinin-3 Coolant Transient Benchmark. The benchmark contains four exercises, but this article shows the results of the exercise two of the VVER-1000 reactor core calculation. Reactor core analysis code SKETCH-N/SKAZKA is used to simulate the steady-state and transient conditions of the VVER reactor. The reactor core boundary conditions for the problem are provided by ATHLET/BIPR-VVER code.In the initial steady-state condition and at the end of the transient the power distribution is compared with the nuclear power plant (NPP) data. The average error in 3-D power distribution is 3.6%. The average difference in fuel assembly (FA) power is less than 2%. In the transient calculation, we compared with the experimental data the reactor power and axial offset. There is close agreement between the numerical results and NPP data.

Performance of ENDF/B-VIII.0 library for VVER reactors criticality safety, fuel depletion and reactor dosimetry applications

The latest ENDF/B nuclear data library released in 2018 is the result of a new international approach to develop evaluated nuclear data for general purpose applications. In order to use the latest ENDF/B-VIII.0 nuclear data library in safety analyses, validation calculations for available benchmark experiments and comparison calculations for previous library versions are required. Validation of this new library for three important reactor areas, criticality safety, fuel depletion and reactor dosimetry, was performed with MCNP6, KENO-VI and Serpent 2 codes for VVER application. Criticality safety and fuel depletion results show better agreement with ICSBEP and SFCOMPO benchmark database cases for the ENDF/B-VIII.0 library when compared with previous libraries including ENDF/B-VII.1. Onthe other hand, using the latest VIII.0 library for VVER reactor dosimetry applications including neutron fluence on reactor pressure vessel and ex-vessel neutron fluence activation monitors is not recommended since the library gives significantly lower neutron fluence and detector activities when comparing to number of alternative nuclear data libraries. Two nuclides that are dominantly responsible for the discrepancies between the latest ENDF/B library and other libraries were identified as O-16 and Fe-56. The conclusions are limited toVVER application where the water gap between the reactor core and the pressure vessel issignificantly decreased compared to PWR.

Investigation of the Cooling of Water-Cooled and -Moderated Reactors Based on Electricity Generation Via Residual Heat in Emergency Situations with De-Energization

Safety enhancement of NPP power units with water-moderated and -cooled reactors on the basis of installation of additional multifunctional low-power steam-turbine generators is studied. It is shown that the parameters of the primary and secondary loops change as the VVER-1000 reactor is cooled via the use of the residual heat in the core to generate steam that serves as the working body for a low-power steam-turbine plant. Uninterrupted power supply for the NPP’s own needs is achieved on account of the fact that in normal and emergency regimes the low-power turbine unit supplies electricity to power consumers during cooldown of a reactor in an emergency situation with a blackout. It is shown for the example of 6 and 12 MW turbine units that the residual heat from one VVER-1000 reactor is sufficient to cool two such reactors for 72 h with de-energization of the NPP and depressurization of the primary loop in one of them.

Investigation of VVER-1200 reactor pressure vessel’s material

At high temperature and pressure, the reactor materials are exposed to various radiation. Irradiation causes significant changes in the crystal structure and their mechanical properties. The mechanical properties of low alloy steels generally do not change so much with temperature. The aim of this research is to prepare the basic measurements to investigate 15Cr2NiMoVA steel, according to MSZ EN ISO 148-1:2017. Charpy impact test was made at different temperatures from -75 °C to 200 °C. Potentiodynamic tests were used to determine the corrosion rate. During the optical microscopy measurement, MnS inclusions were found in the examined material.

Preoperational Tests of the PIRATE Program for Identifying and Assessing the Depth of Defects in Heat Exchangers of Nuclear Power Plants’ Steam Generators

The paper presents the results of acceptance and pilot tests developed by the authors of the PIRATE program designed for automatic analysis of eddy current signals during routine monitoring of the metal state of the heat-exchange tubes in steam generators of nuclear power plants with VVER-440 and VVER-1000 reactor plants. As a result of “blind” acceptance tests using certified standard equipment, the detection of realistic defects on specially prepared tube samples with four gradations of the characteristic size (defect depth) and the reliability of the eddy current testing results with a different level of rejection criterion were determined. The obtained values of detectability and depth of defects satisfy the regulatory documents’ requirements. According to the results of acceptance tests, it is possible to use the PIRATE program at nuclear facilities. The purpose of the pilot tests was to compare the results of operational testing obtained during scheduled preventive repairs by regular systems along with the conclusions of experienced expert analysts to the results of the analysis of the same signals by the PIRATE program in automatic mode. Pilot tests were carried out using data from operational monitoring of steam generators for nuclear power plants with both VVER-440 reactors and VVER-1000 reactors. Findings of defects with varying depths also meet the requirements of industry guidance documents. Based on analyzing the criteria for coincidence of the testing data of the heat-exchange tubes generated by the PIRATE program in automatic mode and a regular system, a conclusion was made about the readiness of the program for industrial operation. In addition, it has been shown that the PIRATE program is no less the foreign programs used at domestic nuclear power plants in some functional capabilities.

Application of Software Tools for Predicting the Corrosion-Erosion Rate to Ensure Integrity of Equipment and Piping of Power Units at Nuclear Power Stations

Basic principles are examined of the application of software tools (ST) for predicting the erosion-corrosion (EC) rate to ensure the integrity of equipment and piping of nuclear power station (NPS) power units. Methods for solving the erosion-corrosion problems encountered in NPS power units have been determined. Software tools that are used most widely throughout the world to solve erosion-corrosion problems in the nuclear power industry are reviewed. The specifics in the application abroad and in Russia are highlighted. Attention is given to the successful foreign experience in improving the corrosion-erosion resistance of piping and equipment based on the solution implemented in designing NPS power units. Information is presented on the development of Russian software tools and the results of their application to render the integrity of piping and equipment of NPS power units with a VVER reactor or fast breeder reactor due to numerical prediction and early detection of the maximum allowable erosion-corrosion metal thinning. The feasibility is substantiated of using the software tools in designing new NPS power units with a VVER-1200 reactor, assessing the residual service life, and prolonging the operation of existing NPSs. Examples are given of the ways for reduction in the erosion-corrosion rate of equipment components and piping of nuclear power units, for example, by controlling the intensity of the hydrodynamic component of erosion-corrosion and proper selection of the metal. It is demonstrated that the cost of power units with a VVER-1200 reactor (including export deliveries) can be reduced by a change-over to less erosion-corrosion resistant but relatively inexpensive steels. The prospects are noted on application in practice of STs to assess the carry over of iron-rich corrosion-erosion products into the working fluid of an NPS with a VVER-reactor. It is pointed out that, when changing over to operation of NPS power units at a super-rated power, this factor should be estimated using STs, and its effect on the wear rate and location of the zones wherein is the highest rate of thinning of the pipeline walls and equipment must be considered.

Computational fluid dynamics analyses of a VVER-1200 nuclear reactor vessel for symmetric inlet, asymmetric inlet, and LOCA conditions

The nuclear reactors can undergo different forms of asymmetric operating conditions with one or more reactor coolant pump (RCP) fails or pressurized thermal shock (PTS) occurrences which can influence the life of reactor vessels. So, accurate descriptions of the heat transfer and the coolant flow within a nuclear reactor for the safety and performance concerns are always desirable. In this paper, computational fluid dynamics (CFD) simulation for a VVER-1200 reactor vessel was presented. Analyses of the reactor vessel were carried out for symmetric and asymmetric inlet boundary conditions, and also for loss-of-coolant accident conditions. Due to computational limits, some simplifications in the geometry of the RPV are introduced. As a result of the analyses, several parameters such as temperature, velocity, and pressure distributions were obtained for the whole reactor vessel. The obtained results for pressure losses through the reactor core matched with the analytical values of the VVER and the predicted thermal results agreed well with the design values. The obtained results show that it is possible to carry out further analysis for several asymmetric or accident conditions utilizing the CFD simulation knowledge obtained.

Optimal design of a VVER-1000 nuclear reactor core with dual cooled annular fuel based on the reactivity temperature coefficients using Thermal hydraulic and neutronic analysis by implementing the genetic algorithms

One of the approaches which can help the enhancement of a reactor power is changing its fuel geometry. For this purpose, as well as decreasing the maximum fuel temperature in PWR reactors, the technology of annular fuels with ability of internal and external cooling shows its importance and has been considered widely. Such fuels are investigated in western PWR and VVER-1000 reactors. Hence, in this study, this fuel in VVER-1000 reactors are considered and studied thoroughly. In this paper, fuel and coolant reactivity temperature coefficients are calculated for a VVER-1000 Nuclear Reactor with Dual Cooled Annular Fuel with changes of internal radius and different power levels and effects of the fuel internal radius on the reactivity temperature coefficients are investigated. By analyzing the fuel internal radius changes in a specific range in the neutronic code, the effects of effective multiplication factor are investigated.For purpose of data fitting, an artificial neural network is trained using the observed data. The input consists of different internal and external radiuses, outputs consist of pitch, fuel and coolant reactivity temperature coefficients. Finally, the optimal geometry of fuel is determined using the neural network by implementing the genetic algorithms based on these dynamic-coefficients. In the optimization process, it has been shown that having an internal radius 2.67 mm and external radius 6.95 mm provides an optimal geometry. Also, validation of the designed artificial neural network and genetic algorithm has been done using neutronic and Thermal hydraulic calculations.

Performance analysis of the second stage accumulators in BNPP-1 in response to MB-LOCA with SBO: A support for practical modification in the injection line

During the last decades, the substitution of passive safety systems instead of active safety systems has increased the reliability of new commercial nuclear power plants. This effort is much more significant for generation (Gen) III + reactors in comparison with Gen II and III reactors. Hydro-accumulators, as a part of the emergency core cooling system (ECCS), is a common passive system among different Gens to encounter the loss of cooling accidents (LOCAs). Recently, the application of advanced and multi-stage accumulators has become successful in GEN III + reactors. Bushehr nuclear power plant (BNPP-1) as a Russian VVER-1000 is somehow categorized between II and III Gens, but it is the first VVER-1000 equipped with two stages of accumulators. This article investigates the influence of Second Stage Accumulators (SSAs) of BNPP-1 on postponing the core damage time in an MB-LOCA with Station Blackout (SBO). Also, SSA's injection line is upgraded after the engineering judgments of start-up tests, and accordingly, each line was equipped with a Flow Restriction Component (FRC). Besides, the effect of FRCs on postponing of the core damage time is assessed in this work. Extracted results achieved from both start-up tests and RELAP5 simulations showed that, although the SSAs are useful in supplying water in MB-LOCA with SBO as a Design Extension Condition (DEC), the application of FRCs are much more effective in postponing the core damage time and without them, SSAs discharge instantly. The results of this study verify the utilization of SSAs with their associated FRCs. Also, the results confirm that some design modifications in BNPP-1 as a commercial reactor can provide resistance against a DECs. Moreover, this experience gained in BNPP-1 could be turned into a basis for safety systems designs in newer VVER-1000s and also VVER-1200 reactors.