Nuclear Power Plant Unit Research Articles

A comparative study of manipulator teleoperation methods for debris retrieval phase in nuclear power plant decommissioning

ABSTRACT This paper presents a comparative study of teleoperation methods of robotic manipulators aiming at application to the decommissioning operation of the primary containment vessel (PCV) of Fukushima Daiichi Nuclear Power Plant Unit 3. A detailed set of mockup is constructed from data collected by internal investigation of the damaged PCV, and a series of teleoperation experiments using a dual-arm robotic manipulator simulating the actual obstacle removal operation is conducted. Three teleoperation methods, manual, teaching-based, and planning-based, are compared in terms of multiple performance indices. The manual teleoperation resulted in low efficiency and safety, and high cognitive load. The method based on pre-recorded teaching data showed the highest safety score, but required large amount of time for preparation. The planning-based method achieved efficient teleoperation with small amount of preparation time, while it was found that it needs further improvement in terms of safety.

Methods for Reducing Short-Circuit Currents in the Generator Circuit When Designing High-Power Units in Nuclear Power Plants

Methods for Reducing Short-Circuit Currents in the Generator Circuit When Designing High-Power Units in Nuclear Power Plants

Результати діяльності Інституту проблем безпеки атомних електростанцій НАН України у 2022 р.

The most important results of the scientific and scientific-organizational activities of the Institute for Safety Problems of Nuclear Power Plants of the National Academy of Sciences of Ukraine (ISP NPP, NAS of Ukraine) in 2022 are presented. It is shown that even under the extremely difficult conditions in which the entire country is due to the invasion of Russian troops into the territory of Ukraine, the employees of the Institute made great efforts to implement planned tasks to ensure nuclear and radiation safety of the destroyed fourth power unit of the Chornobyl nuclear power plant, handling of nuclear hazardous materials, as well as increasing the reliability and safety of nuclear power plants operation. The results of these studies were implemented in the fields of production, electric power industry and during the educational process, as well as highlighted in a number of publications and scientific events. Thus, despite the objective difficulties and guided by the development strategy, the ISP NPP plans to preserve the scientific and technical potential under the conditions of power outages, lack of heating and water supply in some places, disruption of the functioning of public transport, telecommunications and data transmission facilities, etc., as well as to become the leading organization of Ukraine for scientific and technical support of activities for the peaceful use of nuclear energy

Improvement of methods for controlling power oil of cooling tower recycling water supply units at Rivne nuclear power plant

The relevance of the study is conditioned upon the fact that at nuclear power plants, water pumping units using energy oils are operated in the heat exchange equipment of power units. The diagnostic criteria of oils allow identifying defects in the operation of technological equipment. The purpose of the work – to increase the reliability of the operation of oil-filled power equipment by improving the monitoring of the physical and chemical properties of power oil TP-30. The main attention is devoted to increasing the reliability of the operation of oil-filled power equipment by improving the monitoring of the physical and chemical properties of TP-30 power oil. Experimental studies were conducted by chromatography, and gas and liquid extraction using appropriate laboratory equipment. When exploring the content of chemical elements in the segments of the thrust bearing of the cooling tower pumping unit, which is based on Sn, an increase in the content of copper Cu and Sb was observed, which exceeded the standard by an average of 1.2 and 1.1 times, respectively. Most of the analysed physical indicators of oil quality (water content, kinematic viscosity, flash point, acid number) did not demonstrate deviations from the standard values. Only an increase in the mass fraction of mechanical impurities by 0.0026% relative to the standard was noted during the incoming inspection of TP-30 oil. The results of the operational control of the oil in terms of a set of physical indicators fully complied with the established technological standards. The highest content of soluble gases in the oil (0.56% by volume) was recorded for propylene (C3 H6 ). It is recommended to use the relative content of soluble gases in Tp-30 oil to C3 H6 when identifying degradation processes. The absence of residuals of circulating power oil TP-30 in the surface waters of the Styr River during the operational event was established. Generalisations have been generalised about the necessity of expanding the diagnostic criteria for the quality of TP-30 oil, in particular, expanding the list of its physical indicators. In practical terms, the results obtained can be useful for monitoring the quality of other brands of petroleum oils in the systems of turbine units of nuclear power plants, which is important in terms of the safe operation of heat exchange equipment

DEVELOPMENT OF SAFETY ASSESSMENT CODE FOR DECOMMISSIONING WASTE RECYCLING AND DISPOSAL.

In Korea, 10 units of nuclear power plants are projected for decommissioning by 2030, which will consequently generate a large amount of decommissioning waste. It could bring about a lack in the capacity of the radioactive waste disposal facility. In this paper, a computational code, REcycling and DIsposal Safety Assessment has been developed for the safety assessment regarding the recycling and disposal of the decommissioning waste. It is composed of two modules: dose assessment module and maximum allowance concentration estimation module. The dose assessment module is intended to evaluate the exposure doses from radioactive nuclides in the wastes for each recycling or disposal scenario. The maximum allowance concentration estimation module has the function to translate the nuclides concentrations in the wastes from the exposure doses received by the end user for each scenario. The concentration of each nuclide provides the basis for the development of the technical criteria in accordance with the annual dose limit.

Statistical analysis of fuel cycle data from Swedish Pressurized Water Reactors and the impact of simplifying assumptions on simulated nuclide inventories

When analyzing and assessing properties of spent nuclear fuel (SNF) such as radionuclide inventories, the power history of the fuel during its time spent inside the reactor core plays an important role. This information can be very useful in the field of nuclear safeguards wherein a safeguards inspector can use it to verify the fuel properties such as burnup, initial enrichment and cooling time (or collectively termed as the “BIC” set of variables). However, such information may often be unavailable to the safeguards inspector or the level of detail in the available information may be lacking. Therefore, when analyzing SNF for various purposes (such as for safety, safeguards and back-end purposes), the power history of the fuel is most often disregarded altogether and the inspectors only look at the fuel BIC. If the power history-level information is considered, it is not uncommon to make simplifying assumptions about how the fuel is burned in the reactor. In this work, we perform an exploratory analysis of fuel cycle data from two PWR units of the Ringhals nuclear power plant in Sweden. The said analysis describes the variation in the number of cycles, cycle lengths, downtimes et cetera in order to develop a simplified yet representative model of irradiation that may be used to construct synthetic data libraries. Furthermore, we look into impact of changes in the power history on the nuclide inventories of key gamma emitters and isotopes responsible for decay heat in the SNF of the fuel in three different irradiation scenarios. Our results show that in most cases with fuels that are considerably long-cooled, it is acceptable and even preferred to use a simplified power history over an idealized or a representative irradiation model obtained from exploratory analysis of the fuel cycle data. However, for short-cooled fuels, using a simplified or even an idealized history is less preferable over modeling the detailed power history of the fuel due to the presence of short-lived nuclides in the SNF which are more sensitive to variations in the power history.

Assessment of neutron-induced activation of irradiated samples in a research reactor

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

A Pragmatic Approach to Modeling Combinations of Plant Operational States in Multi-Unit Nuclear Power Plant Probabilistic Safety Assessment

One of the technical challenges in multi-unit probabilistic safety assessment (MUPSA) is dealing with numerous combinations of plant operational states (POSs) for each nuclear power plant unit. Since the number of possible combinations of POSs increases exponentially with the number of units, it is impractical to develop separate MUPSA models and assess the risk for each POS combination. This paper proposes a pragmatic approach to modeling combinations of POSs for each reactor unit in MUPSA involving up to 10 reactor units. This approach does not focus on selecting representative POS combinations but rather on screening out non-risk-significant accident sequences in a stepwise manner according to the model quantification results. The effectiveness of the approach is demonstrated by application to cases with four different numbers of units. As a result, in the 2-, 4-, and 6-unit cases, the site and multi-unit core damage frequency (CDF) due to a multi-unit loss of offsite power initiating event are successfully calculated without screening out any accident sequences for each unit. In the 10-unit case, the quantification fails without screening, but it succeeds after reducing the model size by about 43% via the exclusion of the accident sequences in each integrated single-unit model with a CDF contribution of less than 0.1%. The results show that the minimal cut sets obtained in each case cover many POS combinations and that most non-risk-significant POS combinations can be truncated by the cutoff value of 1E-14/yr. In addition, comparing the quantification results according to the stepwise screening criteria shows that the proposed approach can effectively reduce the computational burden in MUPSA without losing much accuracy or realism.

Problems of Main Electrical Equipment Selection in the Generator Circuit of High Power Units of Nuclear Power Plants

Problems of Main Electrical Equipment Selection in the Generator Circuit of High Power Units of Nuclear Power Plants

Research of the Stress-Strain State for Steel Support Structures of Nuclear Power Plant Components under Seismic Loads

Seismic resistance of equipment and piping of nuclear power plant units is determined, particularly, by seismic resistance of their steel support structures. A significant number of these structures are located in the reactor building of nuclear power plants and belong to seismic resistance categories I and II. Therefore, such support structures of nuclear power plant equipment and piping, in general, shall perform their functions under seismic hazards that correspond to the safe shutdown earthquake and design-basis earthquakes, respectively.
 Steel support structures of serial equipment (piping valves, electric and pneumatic drives of piping and pump valves, expansion tanks, etc.) are often designed “on site” in the limited conditions of their actual location. Therefore, it is necessary to design steel support structures of a number of equipment taking into account the fact that it can be connected, in particular, to existing piping with their actual routing. In connection with the above, the research of operation for such support structures under seismic loads is relevant.
 The article presents the research results for the spectrum of natural oscillation frequencies of the equipment support structure, and the stress-strain state of the considered structure under seismic loads is determined. At the same time, the parameters during normal operation and maximum design basis accident at the nuclear power plant power unit are considered. The calculated combination of loads, which simultaneously includes two episodic impacts (maximum design basis accident and an earthquake) is considered.

The New Generation of Moisture Separators-Reheaters for Steam-Turbine Units of Nuclear Power Plants (NPP) with VVER-Reactors

The New Generation of Moisture Separators-Reheaters for Steam-Turbine Units of Nuclear Power Plants (NPP) with VVER-Reactors

Applications of Machine Learning to Consequence Analysis of Hypothetical Accidents at Barakah Nuclear Power Plant Unit 1

The United Arab Emirates (UAE) built four nuclear power plants at the Barakah site to supply 25% of the region’s electricity. Among the Barakah Nuclear Power Plants, (BNPPs), their main objectives are to achieve the highest possible safety for the environment, operators, and community members; quality nuclear reactors and energy; and power production efficiency. To meet these objectives, decision-makers must access large amounts of data in the case of a nuclear accident to prevent the release of radioactive materials. Machine learning offers a feasible solution to propose early warnings and help contain accidents. Thus, our study aimed at developing and testing a machine learning model to classify nuclear accidents using the associated release of radioactive materials. We used Radiological Assessment System for Consequence Analysis (RASCAL) software to estimate the concentration of released radioactive materials in the four seasons of the year 2020. We applied these concentrations as predictors in a classification tree model to classify three types of severe accidents at Unit 1 of BNPPs each season. The average accuracy of the classification models in the four seasons was 97.3% for the training data and 96.5% for the test data, indicating a high efficacy. Thus, the generated classification models can distinguish between the three simulated accidents in any season.

DEVELOPMENT OF NONLINEAR MODEL OF NPP STEAM GENERATOR FOR INFORMATION TECHNOLOGY OF CONTROL OPTIMIZATION

Steam generators of modern power units of nuclear power plants are critical elements of power units and are subject to modernization. Identification of the steam generator model for optimizing the control of the steam generator is an urgent task. The purpose of this article is to develop a nonlinear mathematical model of a steam generator in relative variables for its use in information technology for optimizing control. Mathematical models of heat transfer and steam formation processes in a steam generator are presented in the form of systems of differential equations in relative variables. These models are intended for simulation modeling of thermal processes in a steam generator. Thermal processes are associated with the supply of feed water to the steam generator from the water treatment system and coolant from the nuclear reactor, as well as with the removal of steam from the steam generator to the main steam collector. According to the law of conservation of the momentum of the working medium in the circulation circuit of the steam generator under the evaporation mirror, a nonlinear differential equation for the process of circulation of the steam-water mixture is obtained. A nonlinear differential equation has been developed to calculate the derivative of the steam flow through the evaporation mirror in relative variables. The equations for the auxiliary equipment – the main steam header, the steam turbine valve actuator and the control feed valve actuator – are reduced to relative variables. Using the equations of heat transfer, steam formation, circulation and auxiliary equipment, a nonlinear model of the steam generator in the state space as a control object in relative variables is constructed. Formulas are given for calculating the values of the constant parameters of the steam generator model from the values of its design and technological parameters. A program for a nonlinear mathematical model of the PGV-1000 steam generator has been developed, which is included in the information technology steam generator model module. This will make it possible to solve the problems of identifying and optimizing the information control system of the water level in the steam generator PGV-1000 of the power unit with a nuclear reactor VVER-1000.

Estimation of public exposure during normal operation of unit-1 Barakah Nuclear Power Plant using GALE and HOTSPOT

In August 2020, the first Barakah Nuclear Power Plant (BNPP) unit was connected to the grid and started supplying electricity. The BNPP site consists of 4 pressurized water reactor units, and each unit generates electricity up to 1400 MWe power. Concerning safety, the radiological assessment of the radioactive release and the corresponding public exposure during normal operation is essential. In the present work, the radioactive release during normal operation, in gaseous form, is calculated considering the operating condition of the BNPP by using the GALE code for gaseous forms. Then, the HOTSPOT code is used to simulate the radiological dispersion, including the public exposure and Total Effective Dose Equivalent (TEDE). HOTSPOT code uses the Gaussian dispersion model to provide near-surface releases, short-range dispersion, and short-term releases. Finally, the results in the present work are compared with the values given in the Final Safety Analysis Report of the BNPP. For typical operating circumstances, this study evaluated gaseous source terms such as iodine, noble gases, radioactive particles, tritium, C-14, and Ar-41 and found that TEDE values are within the authorized limits for distances up to 80 kms from the power plant.

TERMIT – Technology of Automated Simulation – an Instrumental Graphic Environment for Developing Digital Twins, Training Simulators, and Simulating Complexes of Nuclear Propulsion Reactor Plants and Nuclear Power Plant Units

TERMIT – Technology of Automated Simulation – an Instrumental Graphic Environment for Developing Digital Twins, Training Simulators, and Simulating Complexes of Nuclear Propulsion Reactor Plants and Nuclear Power Plant Units

Derivation of a new dose constraint applicable to radioactive discharges from Korean nuclear power plants through retrospective dose assessment

A new methodology to derive a dose constraint for radioactive effluent from a unit of nuclear power plant (NPP) through retrospective assessment was developed to reflect operational flexibility in line with international standards. The new dose constraint can retain the safety margin between the offsite dose and the past dose constraints. As case studies, the new approach was applied to 24 Korean NPPs to address the limitations of the existing seven dose constraints that do not fully comply with current international radiation protection standards. Therefore, an effective dose constraint for Korean NPPs was proposed as no less than 0.15 mSv/y, which is comparable to the international practices and previous studies (0.05–0.3 mSv/y). Although the lower bound of the equivalent dose constraint was calculated as 0.17 mSv/y, it is not proposed in this study since the compliance with the derived effective dose constraint can prevent accompanied equivalent doses to any organs from exceeding equivalent dose limits. The new framework and the case studies are expected to contribute toward and support the revision of existing dose constraints for radioactive effluent from NPPs, ensuring better compliance with the current international safety standards as well as reflect the operational flexibility in practice.

Expert express assessment of the impact of heat and mass transfer processes on the residual life of the WWER-1000 reactor vessel due to metal embrittlement

The WWER-1000 reactor is operated at 13 of the 15 operating power units of Ukraine's nuclear power plants (NPPs). Ensuring long-term and safe operation of such reactors is the basis for reliable operation of all 13 Ukrainian nuclear power plants units and the guarantor of Ukraine's energy security. The determining and leading factor influencing the safety and proper residual life of the WWER-1000 reactor vessel is the radiation embrittlement of the reactor steel. The consequences of radiation embrittlement of reactor steel are negatively manifested in emergencies with cooling of the core. This process itself – radiation embrittlement – accumulates constantly and gradually. Therefore, it is important to monitor it by periodically performing ongoing rapid assessments of the brittle strength of the WWER-1000 reactor vessel (along with other factors, including cyclic damage, as discussed in a previous publication). Therefore, it is important to use the calculated express methods of periodic assessment of the brittle strength of the WWER-1000 reactor vessel with guaranteed accuracy. The effectiveness of the approach is supported by low cost of resources – engineering staff, fast and relatively simplified use of computers and software. As an example and confirmation of the applicability of the proposed approach, an expert rapid assessment of the fragile strength and residual life of the reactor vessel of Unit № 1 of the South-Ukrainian Nuclear Power Plant was performed. This takes into account the actual, passport characteristics of its metal. The negative impact of the rigid regime with cooling of the WWER-1000 reactor of Unit № 1, not taken into account by the operating organization (South-Ukrainian Nuclear Power Plant) when extending its designated resource / service life, is shown. timely to clarify complex factors, technical aspects and parameters, as well as – their possible negative effects on the safe operation of systems and elements of nuclear power plants.

The Design of Thermodynamic System Instrument & Control for the Daya Bay Turbine High Pressure Cylinder Retrofit

Before the retrofit, the equipments of the turbine high pressure cylinder of the Daya Bay Nuclear Power Plant is of the old technology. The design of the leaf shape and the steam runner is not advanced, and the equipments are aged after more than 20 years’ operation. From the point of increasing the power and improving the reliability of the equipments, it’s decided to retrofit the turbine high pressure cylinder of Daya Bay Nuclear Power Plant unit 1 and unit 2. The retrofit relates to the design and modification of many secondary loop thermodynamic systems’ instrument & control parameters, in order to match with the first loop reactor control systems, and also the mechanism and electric systems, it’s necessary to verify and optimize these designed parameters of the units after the retrofit. This design and analysis method and experience of the secondary loop thermodynamic systems’ instrument & control parameters can be referred by the similar retrofit of other nuclear power plants.

Simplified Assessment Methodology of Radionuclide Contamination for Subsurface Soil

Determining whether to release a site after decommissioning a nuclear facility should be preceded by an environmental impact assessment of the exposure radiation dose according to the radionuclides in the soil. Currently, in Korea, various evaluation methodologies and decommissioning technologies are being studied for the first decommissioning of nuclear power plants, starting with Kori Nuclear Power Plant Unit (Kori-1), which is based on the “Multi-Agency Radiation Survey and Site Investigation Manual MARSSIM” developed in the United States. The scope and evaluation targets of deep soil may differ depending on the purpose, but it has been confirmed that the International Atomic Energy Agency and the U.S. Nuclear Regulatory Commission are targeting subsurface soil. MARSSIM outlines the need for an evaluation of this subsurface soil but does not suggest specific methods. In NUREG-1757, which complements MARSSIM, it is confirmed that subsurface soil specifically means a soil layer that is 15 to 30 cm deep in the surface layer. In the current study, using the previously verified computational code RESidual RADioactivity (RESRAD)-ONSITE, a methodology for summation is proposed to evaluate the impact of subsurface soil more flexibly and realistically while minimizing the exposure dose evaluation procedure. When using RESRAD-ONSITE according to this evaluation methodology, it was confirmed that it is possible to respond to changes in the depths of various soil layers. In addition, it was also confirmed that this methodology is adaptable to the contamination of nuclides, such as 60Co, 137Cs, 152Eu, and 154Eu, which are expected to be major nuclides when decommissioning nuclear facilities.

An intelligent optimization method of reload core loading pattern and its application

Reload core loading pattern design is the key content related to safety, economy and flexibility of the operation after refueling in nuclear power plant. The optimization design of the reload loading pattern is a very short time requirements, great difficulty, high experience requirements and repeatable work. According to these above rigorous requirements, based on statistical learning and random sampling, this study proposes an intelligent method of reload core loading pattern optimization——Probability Sampling Loading Pattern Optimization (PSLPO). This method intelligently extracts the characteristic law in the optimization space of the reload core loading pattern by statistical learning method to establish the probability model. And the probability model is embedded into PSLPO by the random sampling method. Finally, PSLPO method is constantly updated and evolved in the optimization process by intelligent means such as dynamic adjustment weight factor, probability model adaptive adjustment mechanism and combined simulated annealing design, which realizes the improvement of the optimization ability of the algorithm continuously. According to the actual reload design verification for cycle 5 of Hainan nuclear power plant unit 2, it is shown that the recommended pattern of PSLPO is better than manual search pattern in the aspect of cycle length and power distribution, and the method can intelligently give multiple design patterns, which users can have many choices. Through the optimization evaluation with a single computer core, the optimization efficiency of the method can meet the needs of core reload design. The global optimization ability can be further improved through parallelization and algorithm parameter optimization.