Nuclear Power Plant Unit Research Articles

Analysis of the Opportunity for an Increase in the Thermal Power of Power Generating Units of Nuclear Power Plants (Part 1)

For Ukraine the realization of available reserves to increase the power of operating power units of nuclear plants is a vital problem the solution of which would allow us to increase electric power output. A special role is also played by economic priorities; in particular an increase in power by 1 kW is ten times cheaper in comparison with the construction of 1 kW of new power facilities. One more factor is the world experience in the field of an increase in the thermal power of operating power units of nuclear power plants. Based on the specified urgency this scientific paper analyzes the international experience gained in the field of an increase in the thermal power of the power units for nuclear power plants and describes the main mechanisms and instrumentation required for the realization of the given modification of nuclear power units in Ukraine. The reserves embedded into the project to enable an increase in the thermal power of power units were represented graphically. Recommendations on the scope of appropriate measures to be taken and substantiations required for the implementation of the appropriate level of an increase in the thermal power of Ukrainian power units with water-moderated water-cooled power reactors VVER-1000 and VVER-440 have been given.

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

ABSTRACT Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of 237Np, 238Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of 237Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO2 or UO2–Gd2O3), which was similar to the case of PWR fuel. The differences in the number densities of 235U, 239Pu, 240Pu, 241Pu, 149Sm, and 151Sm have a large impact on k eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.

The Analysis of the System of special water purification of Beloyarskaya Nuclear Power Plant unit BN-800

This article discusses how the latest system of special water purification KPF-30, designed specifically for the fourth power unit of Beloyarskaya nuclear power plant, which has a number of advantages over other water purification systems as chemical-physical and technical-economic, environmental, and other industrial indicators. The scheme covered in this article systems of special water purification involves the use of a hydrocyclone at the preliminary stage of water treatment, as a worthy alternative to ion-exchange filters, which can significantly reduce the volume of toxic waste. The world community implements the project of closing the nuclear fuel cycle, there is a need to improve the reliability of the equipment for safe processes and development of critical and supercritical parameters in the nuclear industry. Essentially, on operated NPP units, the only factor that can cost-effectively optimize to improve the reliability of equipment is the water chemistry. System KPF30 meets the principles and criteria of ecological safety, demonstrating the justification for reagent less method of water treatment on the main stages, in which no formation of toxic wastes, leading to irreversible consequences of environmental pollution and helps to conserve water.

Optimization of controlled processes in combined-cycle plant (new developments and researches)

All modern complex technical systems, including power units of TPP and nuclear power plants, work in the system-forming structure of multifunctional APCS. The development of the modern APCS mathematical support allows bringing the automation degree to the solution of complex optimization problems of equipment heat-mass-exchange processes in real time. The difficulty of efficient management of a binary power unit is related to the need to solve jointly at least three problems. The first problem is related to the physical issues of combined-cycle technologies. The second problem is determined by the criticality of the CCGT operation to changes in the regime and climatic factors. The third problem is related to a precise description of a vector of controlled coordinates of a complex technological object. To obtain a joint solution of this complex of interconnected problems, the methodology of generalized thermodynamic analysis, methods of the theory of automatic control and mathematical modeling are used. In the present report, results of new developments and studies are shown. These results allow improving the principles of process control and the automatic control systems structural synthesis of power units with combined-cycle plants that provide attainable technical and economic efficiency and operational reliability of equipment.

Evaluation of Fast Neutron Fluence for Reactor Pressure Vessel Surveillance of Chashma Nuclear Power Plants Units 1 and 2

The energy-dependent neutron flux incident on surveillance capsule assemblies (SCAs) installed at the outer surface of the core barrel in two pressurized water reactors-Chashma Nuclear Power Plants Units 1 and 2 was evaluated. This safety study was made for the reactor pressure vessel surveillance program to determine the fast-neutron-induced radiation embrittlement of vessel materials in sections facing the core side and the upper weld positions. A total of 68 radiometric sensors were irradiated in each SCA of these two reactors, and dosimetry data were measured from radiometric sensors. Fast neutron flux (E > 1 MeV) was best estimated using the least-squares method by employing LSL-M2 package. Measured results were compared with the fast neutron flux calculated using MCNP5 at the base and weld metal sections of SCA. The results agreed within ±20% of each other in both reactors, which satisfy the regulatory acceptance criteria. Therefore, these results can be used in life assessment of reactors.

Gas transmission services with the combined drive as managed load consumers in regions with the high share of NPP

In this article we consider the questions connected with efficiency of equipment of gas transmission services of gas-compressor stations with the combined drive: with electrical one in addition to the existing gas turbine drive. In our research we consider geological and geographical regional features of the European part of Russia in coordination with an arrangement of large gas trunk pipelines, the existing boosting compressor stations (BCSs), the NPP, high-voltage transmission lines and saliferous basins suitable for creation of underground storage tanks for accumulation of stored air for needs of the gas turbine drive of gas transmission services. The following possible benefits of the schemes offered by authors are researched: the advantage of the additional charge of nuclear power plant units with increase in their capability utilization index during the off-peak periods account of transition to the electrical drive of superchargers; the advantage of use of stored air during the separate periods for partial unloading of the gas turbine drive; replacement of fuel gas as a valuable export resource; the advantage of improvements of the ecological indicators of already operating and expansions of layout opportunities of again constructing BCSs under the terms of acceptable concentration of emissions and to noise indicators. We provided the technique and results of calculation of technical-and-economic efficiency of re-equipment of BCSs under the offered scheme. In the offered method of application are considered: the mean effective performance factor of the NPP, the increase in the capability utilization index due to growth of gross generation and economy of nuclear fuel. We have considered questions of rationalization of configurations of residential settlements adjacent to BCSs under the terms of improvement of ecological indicators and to noise impact of BCSs by various number of working hours of the electrical drive in a year. The offered technique in combination with the specified starting base of data, which bases are provided in this article, will allow to perform the zoned optimization of a profile and the equipment of the compressor yard with BCSs taking into account already existing background and noise pollution Method for calculating the noise level from compressor stations.

Strengthening nuclear reactor safety and analysis

The paper aims at fixing bases for possible strengthening of current Nuclear Reactor Safety (NRS) and safety analysis: this is done by combining the logical frameworks connected with the terms As-Low-As-Reasonably-Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU), Extended-Safety-Margin (E-SM) and Independent-Assessment (IA).ALARA is an early principle in Nuclear Reactor Safety: designers and operators must do their best to minimize doses to the humans. BEPU is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margins constitute an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit, thus forming the E-SM. Here ‘the environment’ includes the personnel in charge of activities connected with the NPP. The E-SM implies the formulation of some ten-thousands safety margins values, which shall correspond to a similar number of monitored variables. IA is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design.The implementation of the idea in the paper brings to an additional safety barrier for existing and future nuclear reactors which may reduce the probability of core melt, presumably at an affordable cost for the industry.

Программные средства, применяемые для анализа и обоснования безопасности атомной электростанции в части нейтронно-физических вычислений

Представлен обзор нейтронно-физических программных средств различных стран, используемых для анализа и обоснования безопасности блока АЭС, в том числе для определения параметров активной зоны блока АЭС в различных режимах эксплуатации, расчета топливной кампании, выгорания ядерного топлива и т. д. Рассмотрены программные средства для создания библиотек малогрупповых констант, включающих нейтронные макроскопические сечения элементов активных зон ядерных реакторов в зависимости от физических характеристик ядерного реактора, программы-имитаторы активных зон ядерных реакторов для сравнительно быстрых вычислений нейтронно-физических характеристик на основе предварительно подготовленной библиотеки малогрупповых констант, и прецизионные программы на основе метода Монте-Карло. Предложены программные средства для проведения нейтронно-физических расчетов ядерных реакторов Белорусской АЭС эксплуатирующей организацией и регулирующим органом и его организациями технической поддержки.

An efficient evaluation of depletion uncertainty for a GBC-32 dry storage cask with PLUS7 fuel assemblies using the Monte Carlo uncertainty sampling method

It is important to evaluate an accurate depletion uncertainty resulting from the limitation of a computer code on accuracy in nuclear depletion calculations from the viewpoint of burnup credit with axial burnup distributions. In this work, the bias and bias uncertainty in keff resulting from the uncertainty in the calculation of isotopic concentrations are assessed using the Monte Carlo sampling method for a generic 32 PWR-assembly burnup credit cask loaded with the PLUS7 assemblies discharged from Hanbit Nuclear Power Plant Unit 3 in South Korea. In addition, an efficient method with the two-way ANOVA is suggested to obtain reasonably accurate values of total depletion uncertainties resulting from the isotopic uncertainties with reduction of Monte Carlo criticality calculations. Finally, the reasonable total depletion uncertainties in keff for 10, 30, and 40 GWD/MTU were efficiently estimated to be 0.011866, 0.013175, and 0.014262, respectively, with 100 Monte Carlo criticality calculations, which have about 1.6, 2.1, and 2.4% discrepancies, respectively, from the ones obtained with 200 calculations. The one for 50 GWD/MTU was efficiently estimated to be 0.030811 with 120 Monte Carlo criticality calculations, which has about 3.6% discrepancy from the one obtained with 200 calculations. Thus, these uncertainties showed good agreements with the uncertainties obtained from 200 nuclear criticality calculations.

Cause Analysis of Pressure Plate Breakage of Valve Limit Switch in Turbine Bypass System of Nuclear Power Plant

The limit switch plates of nuclear power plant unit 1 in the turbine bypass system valve are finding multiple fractures. On the basis of metallographic analysis and vibration analysis, the stress state of the pressure plate is simulated and calculated. The results show that there are some creases in the original plate of the limit switch and the installation error of the pressure plate is the main reason for the break.

Assessment of Allowable Thermal Load for a River Reservoir Subject to Multi-Source Thermal Discharge from Operating and Designed Beloyarsk NPP Units (South Ural, Russian Federation)

This study develops a quantitative approach to the establishment of a maximum allowable thermal discharge into a freshwater body (the Beloyarsk Reservoir) incorporated in the once-through cooling technology at the Beloyarsk nuclear power plant (NPP) (South Urals, Russian Federation). The study is based on a 3-D hydrodynamic model, embracing water circulation, heat transfer, and ice-cover formation in the Beloyarsk Reservoir. The model is driven by atmospheric forcing, river runoff, and the discharge/intake of NPP cooling water. It was used to simulate the horizontal and vertical distribution of water temperature under the effect of the operation of existing (number 3) and anticipated (numbers 4 and 5) nuclear power units. The model is validated by the comparison of the computation results with observed water temperature distribution and ice-cover configuration obtained with remote sensing techniques. The model was also used to predict the future evolution of water temperature after the launching of two new power units, which, having a common cooling system, may affect each other. It was shown that the first of the new units, no. 4, will not dramatically affect the existing thermal conditions in the reservoir, while launching one more unit, no. 5, will apparently result in overheating of the reservoir water in response to the greater volume of cooling-water discharge from the two power units. Because of a specific configuration of the recirculation flow, the reservoir may fail to cope with the dissipation of the generated heat, leading to a steady (uncontrolled) rise of water temperature in the inlet channel to one of the power units. This will reduce the potential of NPP, using the once-through cooling technology, and will most likely have an adverse effect on the survival of aquatic organisms in the Beloyarsk Reservoir. Therefore, some other environment-saving technologies must be developed for removing surplus heat from the unit no. 5 of the Beloyarsk NPP.

Перспективні умови зняття з експлуатації енергоблоків українських АЕС з урахуванням вимог ЄС та МАГАТЕ

Starting from 2018, in different intervals of time, the terms come for completion of the designed operation of power units of nuclear power plants. After that, it is necessary to continue or stop their operation. In any case, the decommissioning of power units of the Ukrainian NPP will take place in the future. Under conditions of Ukraine, the financial support for the decommissioning of power units of the existing nuclear power plants and the management of radioactive waste, which will be formed, is a complex problem. Therefore, the assessment of prospects for the decommissioning of NPP and the financing necessary for this is topical.

Multiple regression approach to predict turbine-generator output for Chinshan nuclear power plant

Abstract The objective of this study is to develop a turbine cycle model using the multiple regression approach to estimate the turbine-generator output for the Chinshan Nuclear Power Plant (NPP). The plant operating data was verified using a linear regression model with a corresponding 95 percnt; confidence interval for the operating data. In this study, the key parameters were selected as inputs for the multiple regression based turbine cycle model. The proposed model was used to estimate the turbine-generator output. The effectiveness of the proposed turbine cycle model was demonstrated by using plant operating data obtained from the Chinshan NPP Unit 2. The results show that this multiple regression based turbine cycle model can be used to accurately estimate the turbine-generator output. In addition, this study also provides an alternative approach with simple and easy features to evaluate the thermal performance for nuclear power plants.

Thermal characteristics of spent activated carbon generated from air cleaning units in korean nuclear power plants

To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

Corrosion of the Second Contour of Power-Generating Units of Nuclear Power Plants with WWÉR-1000 Reactors

We present a survey of experimental investigations of the mechanisms of erosion-corrosion damage to the heat-exchange equipment of power-generating units with WWER-1000 (water-cooled, watermoderated) reactors. By using metallographic methods, we study the in-service defects of different regions. It is shown that the nature of defects in thick-walled structures made of pearlitic steel of the 10GN2MFA type is initially explained by the pitting corrosion caused by the electrochemical interaction of the metal with particles of the sludge, i.e., products of the corrosion-erosion wear of the equipment made of copper alloys. We make a conclusion that the complex interdependent processes of corrosionerosion wear of the equipment are running under the conditions of the second circuit. In their final stage, we observe the development of stress-corrosion cracking. Hence, for the prevention of in-service defects of equipment of the second circuit, it is necessary to remove copper-containing structures.

ОСОБЕННОСТИ ПРОЕКТИРОВАНИЯ ДЕМОНТАЖА ОБОРУДОВАНИЯ БЛОКОВ АТОМНЫХ СТАНЦИЙ С УЧЕТОМ НОРМАТИВНОГО РЕГУЛИРОВАНИЯ ИХ ВЫВОДА ИЗ ЭКСПЛУАТАЦИИ

ОСОБЕННОСТИ ПРОЕКТИРОВАНИЯ ДЕМОНТАЖА ОБОРУДОВАНИЯ БЛОКОВ АТОМНЫХ СТАНЦИЙ С УЧЕТОМ НОРМАТИВНОГО РЕГУЛИРОВАНИЯ ИХ ВЫВОДА ИЗ ЭКСПЛУАТАЦИИ

Thermodynamic evaluation of the solidification phase of molten core–concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

The solidification phases of molten core–concrete under the estimated molten core–concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core–concrete forms (U,Zr)O2 and (Zr,U)SiO4, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO4 requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O2 to tet-(U,Zr)O2, followed by the formation of (Zr,U)SiO4 by reaction with SiO2. Therefore, the formation of (Zr,U)SiO4 is limited under quenching conditions. Other common phases are the oxide phases, CaAl2Si2O8, SiO2, and CaSiO3, and the metallic phases of the Fe–Si and Fe–Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

Feasibility Studies on Pyro-SFR Closed Fuel Cycle and Direct Disposal of Spent Nuclear Fuel in Line with the Latest National Policy and Strategy of Korea

With a view to providing supportive information for the decision-making on the direction of the future nuclear energy systems in Korea (i.e., direct disposal or recycling of spent nuclear fuel) to be made around 2020, quantitative studies on the spent nuclear fuel (SNF) including transuranic elements (TRUs) and a series of economic analyses were conducted. At first, the total isotopic inventory of TRUs in the SNF to be generated from all thirty-six units of nuclear power plants in operation or under planning is estimated based on the Korean government’s official plan for nuclear power development. Secondly, the optimized deployment strategies are proposed considering the minimum number of sodium cooled-fast reactors (SFRs) needed to transmute all TRUs. Finally, direct disposal and Pyro-SFR closed nuclear energy systems were compared using equilibrium economic model and considering reduction of TRUs and electricity generation as benefits. Probabilistic economic analysis shows that the assumed total generation cost for direct disposal and Pyro-SFR closed nuclear energy systems resides within the range of 13.60~33.94 mills/kWh and 11.40~25.91 mills/kWh, respectively. Dominant cost elements and the range of SFR overnight cost which guarantees the economic feasibility of the Pyro-SFR closed nuclear energy system over the direct disposal option were also identified through sensitivity analysis and break-even cost estimation.

Potential and innovative development prospects of Russia’s nuclear industry export

There are considered the key milestones in the development of the Russian nuclear power: the fi rst studies of nuclear physics (the 1920s – 1930s); implementation of the «atomic project» (the 1940s); large-scale development of civil nuclear energy (the 1950s – 1980s); a crisis and restoration of nuclear industry (1990 – 2007); active development on the global nuclear energy market (2007 – present). Analysis is provided of the current state and development prospects in all the key segments of the nuclear industry in Russia: uranium mining, uranium conversion, uranium enrichment, nuclear fuel fabrication, construction of nuclear power plants (NPPs), nuclear power engineering, operation of NPPs and power production, maintenance of NPP power units, decommissioning of NPPs, handling radioactive waste and spent nuclear fuel. Russia is identifi ed as one of the global leaders in all these segments of the world nuclear market (having, among other things, the world’s largest portfolio of NPP orders and the largest uranium enrichment capacities). The export potential for each of the nuclear industry segments under consideration is identifi ed and assessed. It is shown that the segments of NPP construction and manufacturing of nuclear power equipment production are of the largest export potential for Russia. At an early stage of the nuclear fuel cycle, the segment of enriched uranium product and nuclear fuel fabrication has the greatest prospects for growth in export supplies due to the growing number of NPP units of Russian design in the world and Russia’s entry to the market of fuel fabrication for reactors of foreign design.

Dose rate evaluation of workers on the operation floor in Fukushima-Daiichi Unit 3

At Fukushima Daiichi Nuclear Power Plant Unit 3, installation of a fuel handling machine is planned to support the removal of spent fuel. The dose rates at the workplace were calculated based on the source distribution measured using a collimator in order to confirm that the dose rates on the operation floor were within a manageable range. It was confirmed that the accuracy of the source distribution was C/M = 1.0-2.4. These dose rates were then used to plan the work on the operation floor.