Operation Of Nuclear Power Plants Research Articles

Atmospheric 14CH4, 14CO2 and 37Ar measurements around a Swiss pressurized water reactor during an annual revision period

Since the 1980s, radiocarbon (14C) has gained attention as a valuable tool to quantify the amount of fossil and non-fossil emissions of CO2 and CH4 in the atmosphere. Since the 1970s, however, important 14C emissions in the atmosphere also occur through the operation of nuclear power plants. The limited knowledge about these emissions challenges the use of 14C as a universal source apportionment tool. Depending on the reactor type, 14C is emitted in different forms; in particular, pressurized water reactors emit 14C as a mixture of 14CH4 and 14CO2. However, few atmospheric 14C measurements close to nuclear power plants are available, which mostly address 14CO2 emissions. Argon-37 (37Ar) can also be produced in nuclear reactors; however, its atmospheric measurement is challenging, resulting in limited available data. In this study, we sampled ambient air during 20–75 min into 18 individual bags around the pressurized water reactor in Gösgen, Switzerland, at the beginning of the annual revision period in 2019, when 14C and 37Ar emissions can be expected due to the depressurization of the reactor. These samples were analyzed for 14CH4, 14CO2 and partly for 37Ar. About 1 km downwind of the stack, we found background-corrected activities up to 1900, 370, and 93 mBq m−3 respectively. Considering corresponding background activities of 0.3, 48 and 2 mBq m−3 for 14CH4, 14CO2, and 37Ar, this represents an excess of about 6300, 7.4, and 47 times, respectively. Using an atmospheric dispersion model, we satisfactorily simulated the 14CH4 and 14CO2 activities in the surroundings of the reactor during this event. Our measurements emphasize the importance of nuclear power plants in the interpretation of atmospheric 14C measurements and show that pressurized water reactors represent a serious limitation in the use of 14C for source apportionment of CH4 sources. Our results also provide insights into the approximate magnitude of civilian 37Ar emissions from nuclear facilities specifically during maintenance operations.

Aspects of commercialization of small nuclear power plants in Russia and Mongolia

Subject. The article considers prospects for the development of nuclear energy, global problems of energy supply. Objectives. The aim is to identify the socio-economic efficiency of creation and operation of low-power nuclear power plants. Methods. The study employs methods of systems, comparative, structural, and logical analysis. Results. We analyzed prospects for the operation of low-power nuclear power plants, revealed their advantages and disadvantages, proposed recommendations to increase the level of usefulness and reduce the cost of low-power nuclear power plants. Conclusions. Small-scale nuclear power plants have the potential to provide a more sustainable and secure energy future for the world.

The development of international civil liability legal regime for nuclear damage and their inspiration for domestic law-making

As International Atomic Energy Agency has stated in its Handbook on Nuclear Law, “Even in situations for which the highest standard of safety has been achieved, the occurrence of nuclear accidents cannot be completely excluded.” Therefore, the international legal framework for nuclear damage compensation liability has been evolving since the establishment of Nuclear Energy Agency of Organization for Economic Co-operation and Development (OECD NEA) and International Atomic Energy Agency (IAEA). Over the years, various international treaties have been enacted to address the compensation of nuclear damage and to establish liability regimes for nuclear incidents. To date, these treaties have established a series of legal principles of nuclear damage liability, such as the sole liability principle, the strict liability principle, the financial guarantee principle etc., which have been developing since establishment. This paper offers an overview of the historical development of the principles of these international treaties for nuclear damage liability and thus draws upon both primary and secondary sources, including treaties, official documents, academic literature, and reports by international organizations. Including the legislation study methodology, comparative methodology is also adopted in this paper to analyze the changes and trend of these principles. The paper reveals that the Paris Convention, which was established in 1960, was the first attempt to establish a comprehensive legal regime for nuclear damage liability. Most of the principles of this Convention have been inherited by subsequent international treaties and domestic legislations. With the awareness of protecting public’s rights having been significantly strengthened, the range of compensation has been broader, the matters of immunity from liability for operators of nuclear power plants have been reduced, the limitation of the compensation amount has been higher etc. In conclusion, the international legal regime for nuclear damage liability has been showing a shift from protecting the development of the nuclear industry to a joint protection of both public health and rights and the nuclear industry, which should be paid attention to and deeply learnt by domestic legislators of all states for the establishment and perfection of their domestic legislation in this field.

Matching Analysis of Technical Parameters and Safety Standards for Nuclear Replacement of Coal-Fired Units

In the context of the growing share of renewable energy and the impending decommission of a large number of coal-fired units, nuclear energy is the only green energy that can replace coal power for a stable, clean, efficient, and large-scale power supply. This article compares the differences between coal power and nuclear power in terms of thermal system, thermal cycle, turbine parameters, and safety. It discusses the possibility of replacing the boiler of a coal power plant with nuclear power, that is, replacing the boiler of a coal power plant with a nuclear reactor for generation/heating/cogeneration. For coal-fired units with similar capacity that do not use a reheat cycle (at or below high pressure) and nuclear power units (such as a high-temperature gas-cooled reactor), as well as coal-fired units with a reheat cycle (ultra-high pressure and above) and nuclear power units (such as pressurized water reactors), there are great differences in steam parameters. In terms of steam turbines, the size of nuclear power units is relatively larger, requiring additional dehumidification measures. In addition, the safety factors and management methods considered in the site selection, construction, and operation of nuclear power plants are more stringent and complex, and comprehensive analysis and evaluation are needed in aspects such as waste treatment and accident emergency response. Except for the relevant provisions of the American Society of Mechanical Engineers code for pressure vessels, nuclear power units are not compatible with coal-fired units in terms of safety standards. Therefore, considering comprehensively, it is believed that the scheme of nuclear power replacing coal-fired units for power generation/heating/cogeneration program is not feasible at present.

Protection of Nuclear Power Plants from Fraudulent Items

At present a lot of problems with fraudulent items occurred. They disrupt the safety of industry, power engineering and other objects that are important for the life of human society. The article gives infor-mation on causes of occurrence of fraudulent items in nuclear power plants and shows that protection against fraudulent items needs to perform during the whole supply chain of critical items. In detail, it concentrates to the quality of awarding terms of references, which nuclear power plant operator prepares when it gives order to commercial supplier and of process of take-over control of the deliver commodity. Due to national security, the safety of nuclear facilities is also supervised by national nuclear regulatory body. During our research we processed checklists by which national nuclear regulatory body can reveal risks in awarding the contracts to commercial suppliers and in take-over control, and ensures improve-ment of these processes.

Investigation of Abnormal Level Control Oscillations in a BWR Feedwater System

In the long-term operation of nuclear power plants, the aging of systems, structures, and components can lead to maintenance issues that must be dealt with to maintain cost-effective plant operations. One common issue affecting the currently operated boiling water reactors is the onset of unexpected level oscillations in feedwater heaters. This phenomenon can cause excessive cycling of drain valves and lead to premature failures. In this work, we develop a dynamic model of a set of feedwater heaters to determine the root cause of oscillations observed in an operating plant. Simulation results of various transient scenarios were used to investigate the effects of the controller parameters, boundary conditions, and possible valve and instrument issues. The analysis led to the conclusion that the most likely causes of the observed self-sustained oscillations in the system are the nonlinear behaviors of the drain valve and the level transmitter induced by degraded equipment condition. A partial plug of the pressure line used for level sensing in the system can account for a significant deadtime in the level transmitter, a nonlinear effect shown to induce self-sustained oscillatory behaviors.

Methodology for generating wildfire hazard map for safety assessment of off-site power systems against wildfires

Nuclear power plants (NPPs) transfer electricity using off-site power systems. Wildfires can disrupt these systems and cause a loss of off-site power, which can impact the operation of NPPs. In South Korea, wildfire incidents in 2000 and 2022 interrupted the operation of off-site power systems, affecting the NPPs. Furthermore, rising temperatures due to climate change are increasing the frequency of wildfires. Therefore, NPPs are expected to become more vulnerable to wildfires. Wildfire hazard maps are necessary for analyzing the impact of wildfires on NPPs. This paper proposes a methodology for generating a wildfire hazard map via Monte Carlo simulation. The proposed methodology was applied to the vicinity of the Hanul NPP, and a wildfire hazard map was generated.

A Novel Approach for Predicting the Mid–Long-Term Radiation Dose in the Case of a Hypothetical STSBO Nuclear Accident for an Operating Nuclear Power Plant

Four severe nuclear accident scenarios have been identified for operating nuclear power plants (ONPPs). However, there is a research gap in predicting the mid–long-term radiation doses for these scenarios. This study aims to address this gap by proposing a novel approach for predicting the mid–long-term radiation dose in the case of a hypothetical short-term station blackout (STSBO) scenario, one of the aforementioned scenarios. Firstly, the Weather Research and Forecasting (WRF) model was coupled with the Hybrid Single-Particle Lagrangian Integrated Trajectory (HYSPLIT) (WRF-HYSPLIT) model to establish an atmospheric transport and diffusion model for airborne radionuclides, and the regularity of the atmospheric transport and diffusion for the airborne radionuclides was determined. Subsequently, the Residual Radioactive Material Guidelines (RESRAD) OFFSITE (RESRAD-OFFSITE) code was utilized to establish a radiation dose model for predicting the mid–long-term radiation dose resulting from the airborne radionuclides, and the evolution of the mid–long-term radiation dose was analyzed. Finally, the proposed approach was applied to an ONPP, and the results were used to predict the mid–long-term public radiation dose. The results indicated that the total radiation dose would be lower than the dose limit recommended by the International Commission on Radiological Protection (1 mSv/yr) from the second month to the 100th year after the hypothetical STSBO nuclear accident, and the total radiation dose would decrease slowly over time. Recommendations are made for offsite emergency response measures. These research findings can assist ONPPs in analyzing their environmental impacts in the event of an STSBO scenario.

Effect of Constant Cyclic Stress Coupling on the Fatigue Behavior of 304LN Stainless Steel.

The stress state of the primary circuit main pipeline is very complex mixed constant stress and periodic cyclic stress during the operation of nuclear power plants, which often affects the failure behavior of the stainless steel (SS) pipe. The fatigue behavior of 304LN SS under mixed constant stress and cyclic stress was investigated, and it was found that the coupling effect of constant stress and cyclic stress has a significant collaborative acceleration on the fatigue behavior. The research results showed that the cracks in the 304LN SS did not propagate even after 78 days under both constant load conditions of 5 KN and 7.5 KN, while the crack growth rate (CGR) of the specimen increased by about an order of magnitude when coupled with a cyclic fatigue load. The fatigue life of 304LN SS was the longest under 200 °C high-temperature water, and its life was roughly the same under 250 °C and 300 °C water. In a 300 °C high-temperature water environment, the fatigue life under the coupling of constant stress and cyclic fatigue stress is significantly lower than that under symmetric cyclic stress alone. There is a synergistic acceleration effect on the fatigue life of 304LN SS when constant stress is coupled with periodic cyclic stress, which is attributed to the combined effect of a corrosion environment and mechanical factors.

Materials Challenges of the New Nuclear Power Plant in Hungary

Having the experience accumulated during close to twenty thousand reactor years worldwide one may say that enough knowledge is available on the behavior of nuclear structural materials. In recent years however new questions emerge in close association with the long-term operation of nuclear power plants. The questions are in relation with both the operating and the future plants. Hungary operates four Russian designed VVER-440 nuclear units, and currently, two VVERs are under construction. The new, VVER-1200 type reactors represent the generation 3+ which is the latest and the safest version of the world's reactor fleet in operation. Service life of VVER-1200 reactors is 60 years, and operation beyond this term is foreseen. The structural materials of the main and usually non replaceable pressurized components such as reactor pressure vessel, steam generator and so on have to resist load and environment during the long operation period to ensure the components' structural integrity. It is right to say that the long-term, safe operation of the current and future reactors is ultimately governed by the performance of the structural materials in the mechanical technological systems. After introduction of the evolution process of the VVER structural materials performance the article reviews the major loading and environmental parameters of the operation and the ageing effects induced by the operation. Then the most important materials aspects and challenges are presented and discussed.

Seismic fragility analysis of nuclear containment structure with Bayesian cloud analysis framework

Nuclear containment structure plays a critical role in safe operation of nuclear power plant. Seismic safety of nuclear power plant structures has become a hot issue in nuclear engineering community following Fukushima nuclear accident. Seismic fragility analysis, as a key element in probabilistic safety assessment of nuclear power plants, is widely used in nuclear power plant structures. This study presents an innovative method for seismic fragility analysis of engineering structures. The proposed methodology combines conventional Cloud analysis method and Bayesian framework. The accuracy and efficiency of the proposed methodology were demonstrated by a single-degree-of-freedom system, heavy computational efforts but high accuracy seismic fragility analysis method was adopted for comparison. Seismic fragility analysis of a typical prestressed concrete nuclear containment structure was analyzed using the proposed method. Results indicated that method proposed in this study can improve the accuracy of seismic fragility analysis. Furthermore, with the enhance of performance level, the increasing ratios for median seismic capacity of the nuclear containment structure was decreasing.

Remaining useful life prediction of nuclear reactor control rod drive mechanism based on dynamic temporal convolutional network

The control rod drive mechanism (CRDM) is a critical equipment of the nuclear reactor, and the prediction of its remaining useful life (RUL) is important for the efficient maintenance and ensuring the safe, reliable operation of nuclear power plants. In this paper, a novel framework for the RUL prediction of CRDM is proposed, which is a dynamic temporal convolution network (DTCN) based on dynamic activation function and attention mechanism. Firstly, the temporal convolution network (TCN) is used as the backbone of the prediction model, to extract the temporal dependence of the input data. Next, the dynamic activation function DReLU is integrated into the TCN, which can dynamically activate features and capture variable degradation information. Then, introducing the attention mechanism improves the influence of important high-level features extracted by the network on RUL prediction, thereby improving the efficiency of feature extraction in the network. Finally, the DTCN outputs the predicted RUL by performing non-linear mapping on the extracted features. The CRDM accelerated life test platform is established and a series of experiments are conducted using the collected CRDM full-life vibration dataset. The results demonstrated the performance and advantages of the proposed method.

Low-level ionizing radiation-induced DNA responses in the Asian green mussel Perna viridis

Cesium-137 (Cs-137) is a radioactive isotope present in marine environments due to the operation of nuclear power plants and weapons testing. Radiocesium poses a potential risk to marine life due to its long half-life and bioaccumulation. This study evaluated the genotoxicity of low doses of Cs-137 in the Asian green mussel Perna viridis, a sentinel species for marine pollution monitoring, by performing the comet assay and micronucleus test on hemolymph samples. Genotoxicity was assessed after exposing mussels to Cs-137 at dose rates of 0, 5, 10, and 15 μGy/h for 48 h. Cs-137's organ-specific distribution was also determined using HPGe gamma spectrometry. Even at low radiation doses, Cs-137 was found to exert genotoxic effects. Significant increases in DNA strand breaks (%Tail DNA) and micronucleus formation (MNF) were observed at all tested dose rates compared with the levels in controls, with dose-dependent responses. Cs-137 predominantly accumulated in the soft tissues, specifically the gills and digestive gland. The findings support the recommended safety level of 10 μGy/h for aquatic organisms, suggesting its appropriateness as a fundamental criterion for developing the national marine water quality standard for Cs-137 in Thailand.

Assessment of Power Equipment Operational Safety in the Sustainable Management of Residual Lifespan

Introduction. Ensuring the safe operation of energy facilities is a critical issue at all life cycle stages. It is associated with the adoption of management decisions when rescheduling and assessing the remaining resources, which are made based on a comprehensive assessment of the technical condition of the equipment, taking into account the economic efficiency and environmental requirements for further operation, which is the basis for technical and economic assessments. Aim and tasks. This study aims to develop a technical and economic assessment of the safe operation of power equipment to ensure sustainable management of the remaining service lifespan. Results. An algorithm for assessing pipeline system elements by the coolant environment is proposed, which allows for the pipeline load and changes in erosion and corrosion caused by coolant movement. It was proposed that the equipment be systematised by safety class and coolant. When conducting a technical and economic assessment, these parameters and the costs of extending the service life and upgrading the existing equipment were considered in the standard calculations of a certain type of equipment. For this purpose, mathematical models for calculating the residual life for each type, obtained and confirmed based on empirical studies, are proposed. Research findings into ensuring the reliability and safety of operation with various heat carriers have shown that it is necessary to consider low-cycle and high-cycle loads and erosion processes. It was found that the erosive wear rate is 0.3-0.4 mm/year for bending sections of the main circulation pipeline, where maximum loads occur relative to low-cycle and vibration loads, and the stress is 212.5 MPa when assessing the residual life of 7 years. Conclusions. The cost-effectiveness of reassigning the service life is reasonable, as it does not require additional investment in upgrading or replacing equipment subject to annual technical diagnostics. The systematisation of pipelines, improved codifiers, lifespan assessment models, and algorithms for evaluating the remaining life under various loadings enhance the regulatory framework for safe nuclear power plant operations and support technical and economic evaluations for extending their lifespan.

EVALUATION OF TECHNOLOGICAL PLASTICITY OF HIGH BORON CORROSION-RESISTANT STEEL DURING HOT DEFORMATION OF PIPES USED FOR SPENT NUCLEAR FUEL STORAGE

Statement of the problem. The sustainable development of the global nuclear power industry in the future depends on how effectively the problems of radiation safety and nuclear proliferation are addressed. Such problems occur, among other things, at the final stage of the nuclear fuel cycle – spent fuel management. This task is currently on the agenda of both the international community and national governments of countries that develop or intend to develop nuclear energy [1−3]. Today, in connection with ensuring the safety of nuclear energy, boron-containing steels are widely used in the world practice as a material for biological protection and for the manufacture of special equipment parts. This is due to the fact that the B10 isotope in steels provides neutron capture [4]. Hexagonal pipes made of high-boron steel were needed to shield the support cover of the compacted spent fuel storage facility at nuclear power plants. Corrosion-resistant steels alloyed with boron are widely used in the nuclear power industry due to their special nuclear properties. During the operation of nuclear power plants, fuel assemblies that have served their useful life must be stored in special storage facilities using containers – hexagonal tubes made of 04Сr14Ti3B2V steel. To reduce costs during in the construction and operation of NPPs, part of the spent nuclear fuel management should be performed at Ukrainian enterprises. This will provide a significant import substitution effect. To manufacture hexagonal tubes from 04Сr14Ti3B2V steel according to the ingot-hot deformation-profiling-heat treatment scheme, it is necessary to develop the parameters of hot deformation and subsequent technological operations. The use of these pipes in compacted spent fuel storage facilities of NPPs will allow to increase the storage capacity by 2 times, which will give a significant national economic effect Material and methods of research. Material used for the study was steel 04Сr14Ti3B2V (CS-82), smelted using two variants (vacuum-induction – “VI” and vacuum-induction followed by vacuum-arc remelting – (“VD”). Steel grade 04Сr14Ti3B2V belongs to high-alloy, corrosion-resistant ferritic steels with a high boron content of up to 2 %. The method of differential thermal analysis was used to study the phase transformation temperatures in the steel. X-ray diffraction and micro-X-ray spectral analysis were used to assess the phase state. The steel samples were tested for weldability After cooling, the welded samples were subjected to X-ray transmission [5]. Hot twisting tests and mechanical tests at high temperatures were also performed. The macro- and microstructure of the metal was studied. Results. The macro- and microstructure of steel was evaluated depending on the smelting method. The phase composition of the steel was studied. The study of the plastic properties of steel ChS-82 by means of penetration and hot twisting tests showed that the temperature range of maximum plasticity is in a wide range (from 1 025 to 1 150 0C) with a rather low resistance to deformation. During hot deformation in the temperature range of 1 175 0C and above, metal fracture was observed along the grain boundaries at the melting points of the Cr−Fe−B boride phase. The optimal temperature range for hot deformation of steel ChS-82 is 1 000−1 050 ⁰С. Scientific novelty. For corrosion-resistant high boron steel used for spent nuclear fuel storage, the temperature range of hot deformation was selected and the structural and phase composition was estimated, which will allow deforming the metal on the piercing mill and manufacturing pipes.

The protection of nuclear installations in time of armed conflict – Old rules, new challenges

AbstractIn 2022, for the first time in some while, the public around the world was confronted with an armed conflict between states, which directly involved a nuclear facility, specifically a nuclear power plant in operation. Unfortunately, the situation following Russia's armed attack on Ukraine on 24 February 2022 and the acts of war around the Zaporizhzhya nuclear power plant once again draw attention to the need to protect nuclear facilities during armed conflicts. Therefore, this paper reviews the relevant rules of public international law and scrutinizes the norms that have been established through international legislation and soft law mechanisms to protect and guarantee the nuclear safety and security of nuclear installations.

Explainable, Deep Reinforcement Learning–Based Decision Making for Operations and Maintenance

This paper presents research that integrates condition monitoring and prognostics with decision making for nuclear power plant operations and maintenance aimed at reducing lifetime maintenance and repair costs. Additionally, a focal point of this research is to make the decisions explainable to operators, improving the trustworthiness of the decisions from what can be considered a black box model. In this work, we develop and evaluate an explainable, online asset management methodology to help reduce lifetime maintenance and repair costs. Using the latest advancements in condition monitoring, inventory management, deep reinforcement learning, and explainable artificial intelligence methods, we create a predictive maintenance methodology that can optimize the maintenance and spare part management of a repairable nuclear power plant system. To demonstrate these methods, preliminary studies were conducted on a representative maintenance system undergoing a stochastic degradation process that requires repairs or replacement to continue operation. Using deep reinforcement learning, we were able to reduce maintenance spending by approximately 50% compared to optimized, time-based maintenance strategies for the chosen system. A key component of our methodology is the integration of Shapley values to quantify the contribution of various factors to the decision-making process. This addition enhances the explainability and trustworthiness of our decisions, providing operators with transparent and understandable insights into the rationale behind maintenance strategies. The robustness and resiliency of our decision policy against observation noise were also thoroughly evaluated, demonstrating its effectiveness in uncertain operational environments.

Research on Assessment Method for Cognitive Influence Factors of Nuclear Power Plant Operators Based on Fuzzy Analytic Hierarchy Process and Deep Learning of Electroencephalogram Signals

This study explores the factors influencing the cognitive processes of operators in digital nuclear power plants, with a focus on the correlation between these factors and electroencephalogram (EEG) features. Initially, based on expert consultations, seven factors were considered: stress, time, fatigue, procedural complexity, user interface experience, procedural clarity, and efficiency. From these, four were identified as the most crucial for each stage of the cognitive process, highlighting their significant roles in influencing cognitive performance and potentially correlating with distinct EEG characteristics. These were assessed using the fuzzy analytic hierarchy process (FAHP) to determine the weightings of influences across the cognitive stages of monitoring, decision making, and execution. Employing a simulated scenario of a steam generator tube rupture, subjective questionnaires were utilized to gauge participant perceptions of influencer impacts at each stage, calculating human factors fuzzy synthetic values. Concurrently, EEG signals were segmented by operational steps, extracting around 114 features across the time, frequency, and time-frequency domains, which were then dimensionally reduced to 17 principal components via adaptive principal components analysis (APCA). A correlation analysis was performed between the human factors fuzzy synthetic values and the APCA-reduced EEG features of participants. Subsequently, the EEG feature columns of the eight selected participants were used as inputs to construct a transformer-based self-attention network model to evaluate the participants’ human factors fuzzy comprehensive values. The findings confirm the transformer model’s efficacy in assessing these values, evidencing a significant correlation between the EEG features and human factors fuzzy synthetic values. Integrating FAHP with machine learning methodologies, this model proficiently estimated operators’ cognitive states during various cognitive processes, significantly enhancing human-machine interface design and the operational safety and efficiency at nuclear power plants.

Comprehensive assessment of the effectiveness of the hydrogen production and transportation system

Based on the energy development strategy of Russia until 2035, the article substantiates the prospect for the development of nuclear energy, according to which nuclear plants are planned to be forced to unload in the range of up to 50% of the rated power while participating in the regulation of daily unevenness of the electrical load. In addition, nuclear power plants will be involved in primary frequency regulation, which is also associated with the unloading mode of operation of nuclear power plants. All these circumstances force us to look for ways to provide nuclear power plants with base load. Along with pumped storage power plants, the article considers an alternative option in the form of using a hydrogen complex based on the electrolysis production of hydrogen, which satisfies the concept of its carbon-free production. The article presents the global dynamics of the introduction of electrolyzers, and in addition to the increase in the share of electrolysis hydrogen produced at nuclear power plants. The hydrogen complex in this case is a means of providing the nuclear power plant with a base load, which involves the consumption of unclaimed electricity at cost. This allows hydrogen to be produced at a competitive price compared to individual producers on the market. The resulting hydrogen is in great demand as a useful product in a number of industries. The article provides examples of hydrogen consumption in various areas of consumption depending on purity, taking into account the requirements of GOST. A number of consumer industries use hydrogen of a special degree of purity, i.e., more than 99.999% vol. In this regard, additional hydrogen purification on palladium membranes was taken into account. A comprehensive assessment of the efficiency of the hydrogen production system was carried out, taking into account its additional purification and delivery to the consumer in various ways, mastered and in demand in world practice. A comparison has been made of the methods of delivering hydrogen to the consumer in comparison with the production of hydrogen by electrolysis at the point of consumption. A comparison was made using the criterion of net present value for the option of third-party consumers who do not have their own hydrogen production.

Verification and Validation of CFD analysis for 1/7th scale model of APR1400 for CVAP

This paper presents a methodology for analyzing the internal flow of the APR1400 Reactor Vessel Internals (RVI) using Computational Fluid Dynamics (CFD). The Comprehensive Vibration Assessment Program (CVAP) is performed to verify the structural integrity of reactor internals for flow-induced vibrations prior to the commercial operation of nuclear power plants. To classify the APR1400 as a valid prototype, it is necessary to demonstrate through successful CVAP that there are no adverse effects. In this study, CFD is used to calculate the local velocity and flow distribution to generate the forcing functions caused by turbulent flow for analysis programs in the CVAP. The Verification &Validation (V&V) procedure recommended by the American Society of Mechanical Engineers (ASME) is performed to verify the accuracy of CFD based on reliability, and it is compared with tests on a scale model. This paper provides valuable insights into the analysis and V&V process for CVAP using CFD, which can help ensure the safe and reliable operation of nuclear power plants.