VVER-1000 Reactor Research Articles

Developing reference-based correlations for temperature distribution in VVER reactor using gene expression programming and single-heated channel approach

Temperature distribution in different channels of a light-water nuclear reactor is essential for Multiphysics coupling approaches in core calculation. Machine learning algorithms trained by well-structured datasets can be applied for the prediction of temperature profiles in different components of the core including coolant, clad and fuel. In this study, gene expression programming (GEP) as an evolutionary machine learning technique is used for developing proper nondimensionalized correlations to estimate the temperature profiles in the VVER-1000 reactor. The single-heated channel approach based on solving mass, energy, and momentum equations is used as a mathematical model for generating datasets to train the GEP. To determine the most influential parameters on the temperature profiles a two-step sensitivity analysis including Pearson correlation coefficient and one-at-a-time sensitivity measure is performed. According to Wilks’ statistical technique, 59 samples are generated to uniformly distribute the uncertainties in the input parameters. By considering 80 % of the acquired data as a training dataset, three reference-based correlations for coolant, clad and center of fuel are developed. Mean Square Error (MSE) and coefficient of determination are used to measure the error of each approach. The high predictability (i.e., R2 > 0.95) presented in the temperature profiles indicates that the GEP could correlate input parameters successfully. The calculations for relative error show that at some axial directions, the maximum relative error can get to 5 %, while the average relative error is 0.21, 0.23, and 1.71 % respectively for coolant, clad, and fuel temperature profiles.

Justification of Physical Characteristic of Mixed Oxide (MOX) in VVER-1200 Reactor

Justification of Physical Characteristic of Mixed Oxide (MOX) in VVER-1200 Reactor

Uncertainty calculation of thermal hydraulic simulation of DEC-A accident scenario at VVER-1000 NPP using CIAU method

All thermal hydraulic best estimate simulations of transient scenarios at Nuclear Power Plants (NPPs) are subject to uncertainties. One method to evaluate those uncertainties is implemented by the ”Code with the capability of Internal Assessment of Uncertainty” (CIAU-Code), which calculates the uncertainty for the parameters primary system mass, primary system pressure and hot rod temperature by accuracy extrapolation.The present work now presents an extended use of the CIAU method, by deriving the uncertainty of the parameter iodine-131 release to the environment. For the analysis, a steam generator hot header break with the assumption of a stuck open safety relief valve (BRU-A) at the first opening at the VVER-1000 reactor was selected. Such transient constitutes a design extension condition A (DEC-A), characterized by multiple failures of safety systems but with the reactor core remaining intact. The thermal hydraulic best estimate simulation was conducted with RELAP5-3D. Additionally, the Iodine Spiking (IS) phenomenon for the examined accident scenario was simulated using two different empirical IS models (NRC IS Model and BOKU University IS Model). To allow the application of the CIAU method for the IS calculation a theoretical derivation is provided.The results show that for the selected DEC-A scenario a high uncertainty range has to be expected. The analysis allows a conservative estimation of the risk expected in the event of the examined transient sequence.

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

In nuclear reactors, the structural materials of reactor pressure vessel (RPV) are damaged by the radiation that produced by fission reactions. Neutron irradiation damage is one of the most critical factors for the degradation of the reactor pressure vessel (RPV). A displacement per atom (dpa) is a standard measurement for calculating neutron-and gamma-induced radiation damage in materials. The aim of this paper was to evaluate the neutrons radiation damage in the reactor pressure vessel of VVER-1200. In this paper, neutron damage factors, including fast neutron fluxes above 1 MeV, above 0.5 MeV, above 0.1 MeV, the damage rate (dpa/s), and Integrated damage [DPA for one effective full power year (EFPY)] were investigated for VVER-1200/AES-2006 reactor pressure vessel (RPV). The calculations are performed by using NJOY code together with the MCNP6 code at beginning of Cycle (BOC). The cross section library used in present calculations is based on ENDF/B-VII.1 library (ENDF71x). The NJOY code was used for generating DPA cross sections in an ACE formatted file that suitable for using in the MCNP6 code. The MCNP6 code was used for simulating the VVER-1200 reactor core and to calculate neutron DPA rates. MCNP6 simulation of the VVER-1200 reactor core identified the area where the RPV neutron radiation is maximized. The integrated damage for a full effective year was evaluated to be 2.128E-04 DPA/EFPD for the VVER-1200 RPV.

Determination of volume of radioactive waste in activated VVER-1200 reactor structures

The results of calculation studies of the amount of solid high-level, intermediate-level, low-level, very lowlevel radioactive waste (HLW, ILW, LLW, VLLW, respectively) generated during neutron activation of structural materials adjacent to the core of the VVER-1200 reactor plant, depending on the time after the final stopping it. An assessment of the volumes of radioactive waste (RAW) of activation origin generated over 60 years of operation of a VVER-1200 type reactor was carried out on the basis of computational studies of the induced activity of structural and protective materials using reactor and Monte Carlo program codes (SERPENT 2, TVS-M, DYN3D , MCU-PD). Specific activities for each isotope, total specific activities and activities, masses and volumes of activated structural and protective reactor materials for various categories of radioactive waste were calculated after the specified period of operation of VVER-1200 and over time after the final shutdown of the reactor. The obtained dependences of changes in the masses and volumes of activated materials over time after the final shutdown of the reactor make it possible to determine these parameters of materials separately for each category of radioactive waste and in total at any point in time. Thus, the masses of activated materials related to radioactive waste will decrease by 264 tons (from 713 to 449 tons) over 200 years of aging, and the volumes will decrease by 74.3 m3 (from 169 to 94.7 m3). Forecast quantitative data on the volumes and mass of activated reactor structures, depending on the duration of their exposure after the final shutdown of the power unit, will make it possible to develop technical proposals, forecast plans and schedules for organizing a radioactive waste management system during the decommissioning of the VVER-1200 reactor plant, as well as possible technological and design solutions for the disposal system for these radioactive waste.

Investigation of VVER-1000 fuel assembly bowing effect on power distribution during cycle using neutron noise adiabatic approximation

Thermomechanical and radiation loads in the reactor core cause lateral changes in the shape of Fuel Assemblies (FAs). This phenomenon, known as FA bowing, has significant implications for reactor safety and operation. Bowing results in alterations to the spacing between FAs, which in turn leads to neutron flux perturbations affecting power distribution. These perturbations are time-dependent due to variations in boric acid concentration during the cycle. Neutron noise theory is employed to analyze these time-dependent neutron perturbations. In this research, it is investigated the bowing effect during the cycle in a VVER-1000 reactor using the neutron noise adiabatic approximation. It was initially considered a VVER-1000 reactor core to verify the calculations algorithm without bowed FAs. Results had acceptable accuracy in power, initial boric acid, and cycle length. Subsequently, it was explored various modes of FA bowing as neutron noise sources the results revealed that the power noise caused by bowing varies throughout the cycle, as expected. The noise amplitude during the cycle depends on changes in boric acid concentration, burnup, size, and the direction of bowing.

Loading pattern optimization of VVER-1000 reactor core based on the discrete golden eagle optimization algorithm

The main features of the loading pattern optimization (LPO) problem, such as high-dimensionality, multi-modality, and non-linearity, make it difficult to achieve a truly optimal configuration. In recent years, metaheuristic methods have been successfully used to solve this problem. In this research, a discrete golden eagle optimization (DGEO) algorithm has been developed to solve the LPO problem in the first cycle of VVER-1000 reactor core. To evaluate the proposed algorithm, a linear multi-purpose fitness function has been used to improve the safety parameters of the reactor core by obtaining a flatter power distribution during the first cycle, and also to enhance the economic parameters by increasing the cycle length and reducing the cost of fuel recycling. For this purpose, a FORTRAN program has been written to map the DGEO algorithm for the LPO problem using the PMAX and PARCS core calculation code to compute the fitness function in each iteration. To speed up the calculations, parallel computing has been applied in the written program. The results demonstrated that the loading pattern, which is suggested by the DGEO algorithm, enhances all the safety and economic parameters in the fitness function. Thus, the DGEO algorithm is highly reliable for the LPO problems in the VVER 1000 reactor core.

Transient Analysis of LBLOCA with Subsequent Loss of Offsite Power and Loss of All AC Power in VVER-1200 Reactor

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.

An analysis of the “Fast” noise measurements of the dynamic VVER processes

Studies of maneuverable modes of VVER to confirm the possibility of participation of nuclear power plants in the mode of daily carrying capacity have been conducted for quite a long time. Tests at various nuclear power plants with VVER-1000 (Zaporizhzhia NPP in 1998, Khmelnitsky NPP in 2005, Tianwan NPP in 2007) have shown the practical possibility of NPP participation in the daily schedule of carrying capacity, however, the commissioning of nuclear power plants with VVER-1200 requires similar work on all new units with VVER-1200: NVAES-2, LNPP-2 Belarusian NPP. The article presents some aspects of the use of noise control methods for analyzing the condition of equipment and the core. Since the emergence of the technology of noise analysis of signals from VVER reactor installations, researchers have formulated several criteria for obtaining results of appropriate quality. The fundamental requirement for conducting noise experiments was the registration of data in stationary modes of operation of power units, since any non-stationarity made significant changes in spectral estimates, which ultimately complicated the work and “distorted” the results obtained. This requirement was included in the operating instructions of various diagnostic systems using noise signal analysis methods (the SUS system, manufactured by Siemens). For a long period of time, the current situation suited both developers of various diagnostic systems and NPP personnel operating them at power units. On the one hand, this was due to the imperfection of the technical means used (low speed of analog-to-digital converters, limited storage capacity, bulky equipment, etc.), on the other hand, the use of domestic NPP power units only in the base load mode without tracking daily power fluctuations in the power system. The standard archives of the upper block level system, the in-reactor control system and additionally produced multi-channel “fast” measurements with a frequency of 1 kHz for the analysis of maneuverable mode 95-55-95% of the VVER-1200 reactor plant were analyzed. Global disturbances of the core have been detected after one step of the regulatory body of the control and protection system, which attenuates within one second if the next step of the control and protection system has not occurred during this time. Such fast neutron processes can be controlled only by neutron-noise measurements with an upper frequency of at least 20 Hz.

Application of uncertainty and sensitivity analysis in dose assessment during a postulated LBLOCA for VVER-1000 nuclear reactor

A key element in the process of accident analysis for the event of a severe nuclear emergency such as the large break loss of coolant accident (LBLOCA) is to properly quantify safety margins using reliable tools. Assessing the consequences of radioactive material release and the equivalant dose after the LBLOCA can provide clarity about the accident. Reliable safety analysis methods for nuclear power plants (NPPs) can be used to obtain relatively realistic outcomes considering the complex uncertainties of NPP events. In this study, an appropriate technique for sensitivity and uncertainty analysis is employed to develop a methodology for performing dose assessment within a postulated LBLOCA including uncertainties. Considering the approved reliability over the years and safety aspects, VVER reactor is among the most widely used nuclear reactors, which is selected as the case study in this article. The main problem that the study aims to solve is to illustrate whether the inordinate deviation of the input parameters from their reference values lead to violation of the allowed TEDE value by conducting uncertainty analysis. For performing calculations of total effective dose equivalent (TEDE), the environmental contamination with radioactive materials has been simulated using HotSpot code and the obtained results were compared with the solutions of JRODOS program. To obtain the most influential parameters on the obtained TEDE values in different distances from the NPP, an appropriate sensitivity analysis has been carried out, following which the Wilks’ statistical method is used to generate 59 data samples with uncertainty propagation and 95 % confidence. The resulting uncertainty bands demonstrate that the maximum TEDE values of calculated data lie within about 7 to 15 mSv in the postulated LBLOCA scenario, which are below the threshold limit of 50 mSv recommended by International Atomic Energy Agency.

Neutronic examination of the U–Mo accident tolerant fuel for VVER-1200 reactors

In this study, we investigated the possibility of employing accident tolerant fuel (ATF) in VVER-1200/V491 assembly without gadolinium-containing fuel rods using the Monte Carlo code Serpent 1.1.7 with ENDF/B-VII cross-section library. The analysis involves assembly design with reflective boundary conditions. To compare the neutronic performances, U–5Mo, U-7.5Mo, U–10Mo, and U–15Mo fuels were chosen in addition to ordinary UO2 fuel. The concentration of 135Xe, 149Sm, fissile and fertile isotopes with burnup, reactivity feedback with fuel temperature variation, and βeff values were calculated. The results indicate that the fuel cycle length increases by 54.27% for U–5Mo, 32.6% for U-7.5Mo, and 13.8% for U–10Mo, while it decreases by 16.4% for U–15Mo fuel. Additionally, the effect of 95Mo content in natural Mo was investigated by reducing the 95Mo concentration. According to the results, each proposed fuel's fuel cycle length extended when the depletion ratio of 95Mo increased. Additionally, the calculations for reactivity feedback guarantee safe operating conditions for all U-xMo fuels.

Effect of the Inclusion of the Detailed Geometry of the Reactor Core on the Calculated Antineutrino Flux from a VVER-1000 Reactor

Effect of the Inclusion of the Detailed Geometry of the Reactor Core on the Calculated Antineutrino Flux from a VVER-1000 Reactor

Modeling of changes in the nuclide composition of VVER reactor fuel using artificial neural network

The paper seeks to give a computer-based model, using an artificially intelligent technique. This is the adaptive neuro-fuzzy inference system (ANFIS), to predict the concentration changes of certain nuclides in the uranium fuel used for VVER reactors. It uses low-enriched uranium dioxide as a fuel in its solid state. The reactivity in the core is controlled by the control rods. Nuclide concentration changes in the reactor fuel, if not monitored, may cause the unsafe operation of the reactor. Hence, the need for this study. The nuclides considered in this study are, U-235, U-236, U-237, U-238, Pu-239, Pu-240, Pu-241, Am-242 and Am-243. The initial computational technique was performed using MATLAB Simulink. The simulation data for all the concentrations of the nuclides were obtained. Then the proposed ANFIS model was performed and tested using data from the Simulink. Results from the simulink and ANFIS were compared and the results were in good agreement. Again, the results were compared to the Calculating Actinide Inventory (CAIN) code from the IAEA-TECHDOC-1535 published in 2007 and both showed a good agreement. An RMSE of about 0.98% and 1.25% were obtained for training and testing data respectively. The developed model will allow technologists to quickly perform calculations for the reactor, which is essential for safety systems. It could be concluded that the ANFIS model can effectively be used to predict the concentration of each nuclide in the uranium fuel because it is effective, precise with lesser error, and does not consume time.

Neutronic Analysis and Fuel Cycle Parameters of Transuranic-UO2 Fueled Dual Cooled VVER-1000 Assembly

Abstract The feasibility of transuranic (TRU) fuel rods has been studied for a dual-cooled VVER-1000 assembly along with the conventional UO2 fuel rods. It has been found that the heterogeneous arrangements of TRU and UO2 fuel rods help to improve the multiplication factor, enhance the fuel cycle parameters, and maintain the negative Doppler coefficient of the reference VVER-1000 assembly. Within the different heterogeneous combinations of TRU and UO2 fuel rods, the equal number of UO2 and TRU fuel rods, which is referred to as model 6 increased the multiplication factor of the reference assembly by 7.66% at the beginning of the cycle. Furthermore, the fuel cycle parameter becomes almost double for this model in comparison with the reference assembly. TRU rods make the neutron spectrum of reference assembly slightly harder and increase the fast fission rate. However, the Doppler coefficient becomes less negative, and flux level decreases with increasing the number of TRU rods. By considering the safety parameters and neutronic behavior, model 6 (equal number of UO2 and TRU fuel rods) is to be effective for the annular fueled VVER-1000 reactor.

Time-dependent leakage model for the identification of defective fuel assemblies of VVER-type nuclear reactors, part 1: Theory

Level of radioactive contamination in primary coolant of nuclear reactors must be kept as low as possible, therefore essential conditions is the leakage-free state of the fuel rods. To monitor the leakage defects in VVER reactors, the radioactivity of fission products must be measured in the primary loops. From radioactivity data it is possible to infer the fuel assemblies' hermetic or time-dependent leakage state. A new, time-dependent leakage model was developed for operating conditions of VVER nuclear reactors. This leakage model is able to calculate isotopes’ radioactivity inside the fuel and in the primary coolant originating from nuclear fission and neutron-activated reactions. The core of the leakage model consists of several systems of differential equations describing dominant nuclear processes on time-dependent number of radioactive isotope nuclei within fuel assemblies and primary coolant. The model considers the effect of technological parameters on the activity of radioisotopes in the primary coolant.

Investigation of prototypical corium interaction with samples of VVER-1000 in-vessel and ex-vessel structures

After the Fukushima Daiichi accident, many countries worldwide conducted stress tests for nuclear power plants (NPPs) in operation. These tests initiated back-fitting activities. One of the long-term back-fitting tasks for VVER-1000 reactors involves managing the ex-vessel phase of a hypothetical severe accident (SA). The challenge is to cool down the corium ejected into the reactor cavity after a reactor pressure vessel (RPV) failure by spreading it onto the dedicated area. However, to transport the corium to this area, the thermal shielding must withstand the load of poured corium. The thermal shielding ability must be investigated to maintain this strategy successfully. Moreover, the internal RPV parts may prolong the time of RPV failure and decrease the temperature of the corium ejected from the failed RPV. This phenomenon needs to be verified experimentally. Therefore, the interaction of simulated corium with the internal RPV parts and thermal shielding samples was studied experimentally. These experimental findings may help to understand the corium behavior and its cooling during severe accidents in VVER-1000 reactors.

Подход к оценке мониторинговых уровней оперативного вмешательства после аварии на АЭС с использованием системы RODOS

The paper presents a comprehensive analysis of the IAEA's methodology for calculating the operational intervention levels (OILs) to protect the public in the aftermath of a radiation disaster based on monitoring data. Developing the OILs assessment approach involves outlining specific indicators based on accident scenario characteristics, geographical location of nuclear power plants, timing after releases, and dose criteria established by the Russian regulatory documents. The paper highlights the feasibility of using a unified computation tool to solve interconnected tasks, such as "prediction of radiological consequences" and "development of monitoring OILs". A methodological approach to assessing monitoring OILs using RODOS models, designed for predicting the consequences of a radiation accident, has been developed. This approach includes three stages: assessing the radiological significance of a scenario, evaluating monitoring OILs for different generic criteria, and justifying the recommended values of OILs. Calculations were performed to show how the new methodology can be used based on the data for the most severe possible accidents at VVER-440, VVER-1000, VVER-1200, and BN-800 reactors. The results showed that the maximum radiation doses for the prospective BN-800 and VVER-1200 reactors do not exceed the dose criteria for evacuation of the public. The dynamics of OILs were computed for the most severe accident scenario at the VVER-440 reactor, using the "ambient dose rate" metric and "evacuation" dose criteria for effective and thyroid exposure, which are justified in the Russian NRB-99/2009. The OILs obtained through the presented methodology have been compared to the aggregated values of the same indicator calculated using the IAEA methodology in a "ground scenario". Further research is needed to validate monitoring techniques for potential accidents at Russian nuclear power plants. The research should consider regional conditions and the Russian regulatory framework.

Радиоэкологическая оценка реакторных установок на основе сценариев проектных и запроектных аварий

The methodical approach based on scenarios of design and beyond design basis accidents was developed for comparative radioecological assessment of reactor facilities. The dynamics of possible radiation doses to the thyroid and the whole body of the public resided nearby the Russian nuclear power plants equipped with the VVER-440, VVER-1000, VVER-1200 and BN-800 reactors was estimated. With the use of the conservative approach for each accident scenario, as well as with the use of acute and medium-term dose criteria, radiation exposure indices were calculated for each accident scenario. Reactor facilities are ranked according to an index summarizing the accident scenarios features and the probability of their realization, the height of radioactive emissions and the activity of dosage-forming radionuclides. The potential radioecological danger of Russian reactors to the population has been reduced. The lowest estimates of the summarized emergency risk are typical for the advanced VVER-1200 reactor of the 3+ generation and the fast neutron reactor BN-800. The most severe emergency scenarios for each considered reactor are highlighted. The VVER-440 reactor has the highest number of severe projected accidents exceeding the dosage criteria envisaging evacuation of the population. The directions of further research related to the analysis of the environmental radiological effects, the justification of radiation protection measures and the development of operational intervention levels in the acute post-accident period are outlined.

Fuel Performance Comparison of Uranium Nitride and Uranium Carbide in VVER-1200 using OpenMC

Nuclear power is a reliable and large-scale source of GHG-free electricity. This study asses the viability of ATF fuel of uranium nitride (UN) and uranium carbide (UC) as fuel for the VVER-1200 reactor. A comprehensive overview of the VVER-1200 and Accident Tolerant fuels is conducted. A review of the development of ATFs identified UN and UC as viable fuels for the VVER reactor. The study utilizes OpenMC to model the VVER-1200 core and compares the behaviour of ATF with conventional fuel. Key findings include comparable k-eff values implying similar neutronic behaviour. UO2 and UC showed similar fission rates across the core while UN showed higher neutron flux and fission rate in the outer part of the core. The base Z44B2 showed increased flux and fission rate with UN as the fuel. ATF behaviour showed to be comparable to the UO2 and thus is a potential alternative to conventional fuels. ATFs provide an additional level of safety because of higher melting points and higher thermal conductivity. This study can be further improved to investigate the depletion of ATFs so that the behaviours of the core over large periods of time, fission products and operator safety can be assessed. Base case k-eff value of 1.24795 are comparable to k-eff values generated by UN and UC.

The effect of subcooled water parameters on thermal-hydraulic characteristics for VVER reactor

The effect of subcooled water parameters on thermal-hydraulic characteristics for VVER reactor