Reactor Plant Research Articles

Polycotton waste textile recycling by sequential hydrolysis and glycolysis

As a result of the current high throughput of the fast fashion collections and the concomitant decrease in product lifetime, we are facing enormous amounts of textile waste. Since textiles are often a blend of multiple fibers (predominantly cotton and polyester) and contain various different components, proper waste management and recycling are challenging. Here, we describe a high-yield process for the sequential chemical recycling of cotton and polyester from mixed waste textiles. The utilization of 43 wt% hydrochloric acid for the acid hydrolysis of polycotton (44/56 cotton/polyester, room temperature, 24 h) results in a 75% molar glucose yield from the cotton fraction, whereafter the hydrolysate solution is easily separated from the solid polyester residue. The reaction is scalable, as similar results are obtained for experiments performed at 1 mL, 0.1, and 1.0 L and even in a 230 L pilot plant reactor, where mixed postconsumer polycotton waste textile is successfully recycled. The residual polyester is successfully converted via glycolysis to bis(2-hydroxyethyl) terephthalate in 78% isolated yield (>98% purity).

Effect of variable angular piping bends geometry on turbulent flow of supercritical water in an SCWR Plant piping network: a 3-D CFD investigation

Abstract One of the most critical components of Super Critical Water Reactor (SCWR) plant is pipe bend, carrying massive volume of supercritical water. Because of its intense fluctuation of velocity and pressure, it can exhibit unusual flow pattern in pipe bends. This unusual flow pattern may finally destroy the integrity of the piping network. Alternatively, because of unique chemical and physical properties, transportation of SCW can adversely effect on the integrity of the pipe bends, indicating probable failure of piping network. Therefore, intensive studies of SCW flow in bends is a must while designing an efficient plant piping network. This article presents CFD analysis, carried out to examine the effect of variable pipe bend angle (90, 120 & 135;) on turbulent flow of SCW. The analysis indicates, the portion of the straight inlet pipe with fully developed flow shows no variation of results in all bend configurations. The study also reveals that secondary currents strengthen at lower bend angles and culminate the formation of vertices at the outlet of the right-angled elbow. The study shows that a greater part of the cross-sectional area is covered by the vortices at the bend mid-planes compared to the bend outlets, where mixing of stream wise flow occurs as well. It is found that the radial positions of each vortex center are closer to the circumference of the pipe, and appear to be insensitive to bend angle. The study is ended with the comparison between supercritical and normal water at standard conditions.

Variation in bioelectricity production in integrated CW-MFC: An insight into coliform inactivation affected by HRT and power density

In the current study, domestic wastewater was treated in three identical vertical up-flow reactors; R1 (constructed wetland), R2 (CW-MFC), and R3 (unplanted CW-MFC) under different HRTs of 36 h (Phase 1), 24 h (Phase 2), and 18h (Phase 3). Periodic reduction of HRT from Phase 1 to Phase 3 resulted in deteriorated organics and fecal coliform removal by the reactors. R2 showed higher pollutant removal and voltage generation in every phase of the study compared to R1 and R3. R2 exerted 93%, 87%, and 57% mean COD removal during Phase 1, Phase 2, and Phase 3; with maximum open circuit voltage generated as 925 mV, 695 mV, and 429 mV respectively. Linear regression analysis showed that operating voltage and power density had significant effects on the variance of effluent fecal coliform concentration. The regression analysis also revealed that 36 h – 24 h HRT was critical where power density influenced the pathogen inactivation considerably. Multiple batch studies revealed the main role of reactor media and plant roots was to support the attached microbial growth for biodegradation of the organics Radial oxygen loss did not affect the anaerobic environment at the anode after 800 days of reactor operation.

Model for integral codes for calculating containment thermohydraulics in severe accidents at nuclear power plants with a water coolant

The article describes a model for calculating the containment thermal in severe accidents at nuclear power plants with a water coolant. The model is implemented in the SOCRAT integral code in the form of a CONT_TH module, which allows performing self-consistent calculation of parameters in a VVER reactor plant and a containment during a severe accident. In conjunction with other modules of the SOCRAT code, a realistic calculation of the transfer of radioactive substances to the containment is provided, which is necessary to calculate source term into the environment when justifying the safety of nuclear power plants, including probability safety assessment level 2. With a slight adaptation, the model can be used to calculate the parameters of containment in a water-cooled NPP as part of other integral codes.

Evaluating the Radiation Characteristics of the Metal Structures of RBMK-1000 Reactor Plants Shut Down for Decommissioning

Evaluating the Radiation Characteristics of the Metal Structures of RBMK-1000 Reactor Plants Shut Down for Decommissioning

Investigation of neutron irradiated W/CuCrZr joints

This study investigates the effects of neutron irradiation on tungsten (W) and copper-chromium-zirconium (CuCrZr) joints under conditions mimicking the high neutron flux environment of a tokamak fusion reactor. Samples of W/CuCrZr joints were subjected to irradiation in the Belgian Reactor 2 (BR2) nuclear reactor at SCK CEN (Belgian Nuclear Research Centre) to simulate the intense neutron exposure characteristic for International Thermonuclear Experimental Reactor (ITER) and DEMOnstration power plant reactor (DEMO) operations. The primary objective was to evaluate changes in the mechanical properties and microstructure of these materials, which are critical for their potential use in plasma-facing components.It is revealed that a significant reduction in tensile elongation of the joint, indicating some degree of embrittlement, is observed after the irradiation. Importantly, this effect is independent of the irradiation temperature. Possible physical reasons for the observed phenomenon are discussed.

Contribution of transient heat transfer to overpressure protection in a passive Boiling Water Reactor

The BWRX-300 reactor is a small modular reactor with 300 MW nominal electric power output. As the abbreviation indicates, the reactor is a boiling type reactor using water as the coolant. The main operating principle of the reactor model is the use of natural circulation of the coolant for both steady-state operation and emergency cooling of the reactor. Isolation Condenser Systems (ICS) have been used in early BWRs and were reintroduced in the 1990s to implement passive safety.Modern technologies enable effective simulation of objects of various complexity. In this study, the thermal-hydraulic system code TRACE v.5 is used to simulate BWRX-300 reactor plant operation. The software enables calculation of a system consisting of the reactor pressure vessel and ICS in both steady-state and transient operation.The ICS is an emergency cooling system borrowed from the previous generation of GE Hitachi BWRs, the ESBWRs, and it consists of a vertical tube bundle heat exchanger immersed in a water tank. Although this system design remains unchanged, its purpose is different. The BWRX-300 design completely eliminates safety and relief valves. This is why the ICS must perform both overpressure protection and long term heat dissipation functions.Three reactor plant operation states were simulated: normal operation with nominal parameters; reactor isolation; and the transient process into the cooling down mode by reactor shutdown and ICS activation. The results demonstrate that the ICS is capable to mitigate overpressure transients. Upon ICS startup, its metallic structures absorb heat well in excess of its nominal steady-state rating. Initial thermal inertia of the system is important for overpressure mitigation in transients.

Design and application of an automated video inspection system for the inner wall of pressure vessels in nuclear power plant reactors

Abstract The pressure vessel of a nuclear power plant reactor is one of the key equipment in pressurized water reactor nuclear power plants, and its inner wall is an essential component of the pressure boundary in the primary loop. The inner wall of the reactor pressure vessel is subjected to long-term alternating stress and fluid erosion caused by the flow of the working fluid, which can easily form various surface defects such as pitting and unevenness. In order to timely detect and handle such defects while taking into account the maintenance period and personnel radiation risks, this paper designs an automated video inspection system for the inner wall of the pressure vessel of a nuclear power plant reactor with telescopic and pitching functions, which realizes the automatic video scanning of the inner wall of the pressure vessel and the precise positioning of defects.

Testing the thermal performance of water cooling towers

Abstract The essential service water system (ESWS) in nuclear power plants circulates the water that cools fan coolers, cooling water heat exchangers, and condensers and other components before dissipating the heat into the environment. Because this includes cooling the systems that remove decay heat from both the primary system and the spent fuel rod cooling ponds, the ESWS is a safety-critical system. In locations without a large body of water in which to dissipate the heat, water is recirculated via a cooling tower. Cooling towers fall into two main categories: Natural draft and Mechanical draft. Since, the mechanical draft cooling towers are much more widely used; the focus is on them in this work to measure its performance. Therefore; the present work addresses the thermal acceptance test for mechanical draft water cooling tower in order to determine its thermal capability in accordance with the cooling technology institute (CTI) code recommendations. Both the Characteristic Curve and the Performance Curve Methods are used. A detailed description of the test procedure according to the CTI ATC-105 test code is presented, and the test code is applied on the Egypt second research reactor (ETRR-2) cooling tower as a case study. All the instruments used in the reactor plant are calibrated prior the test and the measured data is collected at regular intervals according to the code recommendations. The test result shows a tower capability of about 105 % and so the tower is thermally accepted.

Analysis of the application of natural circulation of cooling media in shipboard reactor installations

In projects of floating facilities with a nuclear power plant, great importance is attached to the formation of a set of properties of the internal self-protection of the reactor, i. e. the development of properties that ensure its safety based on natural feedbacks, processes and characteristics. To do this, the RU project is based on physical patterns and technical solutions that prevent or minimize possible consequences from emergency situations. In particular, the natural circulation of the coolant in case of loss of power supply by the circulation pumps of the first round allows cooling the core after its transfer to a subcritical state. However, in order for a natural circulation to occur sufficient to ensure nuclear and radiation safety, certain conditions must be met in a particular case. The paper notes that the most important property of reactor installations of nuclear vessels is their ability to ensure natural circulation of the coolant in emergency situations, as well as when operating at capacity. This makes it possible to reduce the dependence of reactor plant safety on the technical condition of power supplies. Currently, natural circulation of auxiliary cooling media in passive safety systems is also widely used in reactor installations. These include emergency reactor cooling systems using an emergency cooling system, including using intermediate heat exchangers; reducing emergency pressure in the protective shell in case of an accident with a large coolant leak; cooling the reactor vessel in an out-of-design accident with the formation of corium. The paper proposes a method for preliminary analysis of the reactor plant’s ability to ensure natural circulation of the coolant, which allows us to compare various design solutions that contribute to an increase in coolant consumption during natural circulation. It is shown that for the operation of the reactor plant in stand-down mode and at energy power levels with natural circulation of the coolant, it is advisable to switch to boiling reactors. Passive safety systems using natural circulation have received special development on new nuclear vessels of projects 20870, 22220, 10510. Further development of safety systems using natural circulation to move cooling media is predicted.

Probabilistic safety analysis of the large primary circuit leakages accident scenarios in the VVER-reactor

Probabilistic safety analysis of the loss of coolant accidents in the VVER-type reactor plant has been performed taking into account the internal initiating events of the reactor primary circuit large leakages. Swedish program code RiskSpectrum PSA was used for probabilistic safety analysis. Logical probabilistic models of the accident scenarios of the large primary circuit leakages in the VVER-type reactor were developed taking into account different operation modes of the power plant unit. Critical paths and probabilities of the accident scenarios occurrence of the large leakages were identified. The critical path of development of the reactor primary circuit large leakages accident with large leakages of 140–346 mm diameters is the accident sequence with a leak in the pipeline to the reactor pressurizer vessel. It has been established that the greatest contribution to the failure of safety functions during this initiating event is made by the failures due to common causes of the supporting systems (cooling and ventilation systems), critical consumers cooling circuit, emergency injection system. The critical path of the accident with large leakages of 346–850 mm diameters is the accident sequence with the rupture of any of the four primary circuit loops. The greatest contribution to the failure of safety functions during this initiating event is made by the failures due to common causes of the emergency injection system elements. Based on the accident analysis, recommendations for increase of performance reliability of safety functions during the large leakages accidents under all operation states of the nuclear power plant unit were given. In order to increase reliability of safety systems, it is necessary to eliminate failures due to common causes of equipment, increase the reliability of operation of supporting systems, change the maintenance and checking equipment procedures.

Analytic network process (ANP) based decision support tool for nuclear power plant location and reactor type selection

The nuclear energy industry has seen a resurgence due to interest in sustainable energy. Regulatory agencies are evaluating innovative technology and facility designs for the first time in a long time. The nuclear energy business needs a workable instrument for decision-making in light of the shift in public and regulatory opinion. This study has two main motivations. The first is to determine the set of criteria to be used in determining suitable locations for a nuclear power plant (NPP) to be established in Turkey and to determine suitable regions with Analytic Network Process (ANP) based analyzes in the light of these criteria. The second is to select the most suitable nuclear reactor type for the specified location. Plant efficiency and lifetime vary depending on the reactor type used. For this reason, in the study, both the location of the nuclear power plant and the selection of the appropriate reactor type for the plant were evaluated together. In this context, a criterion set consisting of 3 main criteria and 20 sub-criteria to be used in NPP installation location and reactor selection was determined. In the study, site selection was made by focusing on economic, social and technical factors, and reactor selection was made by focusing on technical factors. Giresun, Trabzon, Rize and Artvin provinces located in the Eastern Black Sea region were evaluated for NPP installation. For reactor selection, we focused on reactor types according to their coolants and in this context, PWR, PHWR, GCR and BWR reactors were examined. NPP location and reactor selection in the study; It was realized as Giresun (PWR) > Rize (PWR) > Trabzon (BWR) > Artvin (BWR).

Reactor power control system design and dynamic simulation for load maneuvers over full range and entire lifetime of an integrated supercritical CO2 reactor plant

Due to advantage of high efficiency, low carbon, and flexible operation, the integrated supercritical CO2 reactor plant has promising prospect in energy supply. This paper aims to clarify performance of the reactor plant and establish a reactor power control strategy through numerical modeling. Firstly, performance on separated components is evaluated to provide guidance for control strategy development. Secondly, influence of control methods, controller schemes, and saturation models on control strategy are discussed. Thirdly, dynamic response of reactor plant with proposed control strategy is confirmed via load change transients. Results show that passive control method is infeasible at EOL and load demand below 50 %FP, and active control method must be considered. The proposed control strategy adopts coupled temperature-channel and power-channel scheme and soft saturation model. It ensures safe operation of the plant over full load range in whole lifetime. These results can provide helpful guidance on part-load operation for this system.

Blistering Behavior of Beryllium and Beryllium Alloy under High-Dose Helium Ion Irradiation.

Beryllium (Be) has been selected as the solid neutron multiplier material for a tritium breeding blanket module in ITER, which is also the primary option of the Chinese TBM program. But the irradiation swelling of beryllium is severe under high temperature, high irradiation damage and high doses of transmutation-induced helium. Advanced neutron multipliers with high stability at high temperature are desired for the demonstration power plant (DEMO) reactors and the China Fusion Engineering Test Reactor (CFETR). Beryllium alloys mainly composed of Be12M (M is W or Ti) phase were fabricated by HIP, which has a high melting point and high beryllium content. Beryllium and beryllide (Be12Ti and Be12W) samples were irradiated by helium ion with 30 keV and 1 × 1018 cm-2 at RT. The microstructures of Be, Be12Ti and Be12W samples were analyzed by SEM and TEM before and after ion irradiation. The average size of the first blistering on the surface of Be-W alloy is about 0.8 μm, and that of secondary blistering is about 79 nm. The surface blistering rates of the beryllium and beryllide samples were also compared. These results may provide a preliminary experimental basis for evaluating the irradiation swelling resistance of beryllium alloy.

Risk-based method for substantiating reduction of pressure of hydraulic strength tests and its application on steam line of nuclear power plant with WWER (PWR) reactor plant

Hydraulic tests are used in various sectors of the national economy to check the strength and tightness of vessels and pipelines, their parts and assembly units loaded with pressure. The hydraulic test pressure is a calculated value and is determined on the basis of regulatory documentation for the relevant control objects. Analysis of current trends in the field of hydraulic testing both in the Russian Federation and abroad has shown a trend to reduce the damageability of test objects by choosing the most optimal hydraulic test pressure. The main purpose of the research was to reduce the operational damage to equipment and pipelines of nuclear power plants by reducing the pressure of hydraulic strength tests during maintenance. The research is based on probability theory, mathematical statistics, strength physics and fracture mechanics, taking into account the possibility of occasional discontinuities or defects that might be observed in operation. The method is based on a risk-based approach, which makes it possible to determine the permissible risk variation based on recommendations on the safety of nuclear power plants in the Russian Federation. The objects of research are the equipment and pipelines of nuclear power plants. The result of the research is the development of a method substantiating the reduction of the pressure of hydraulic strength tests for equipment and pipelines of nuclear power plants based on a risk-based approach. The developed method can be used by organizations involved in the production of electricity (at nuclear or thermal power plants), transportation of oil, gas or other substances, at facilities where pipelines are an essential part of their structures, affecting the reliability and efficiency of operation of these facilities.

Computational Studies of Thermal Hydraulics in a New Integral Reactor Plant VVER-I with Natural Circulation

Computational Studies of Thermal Hydraulics in a New Integral Reactor Plant VVER-I with Natural Circulation

Mechanochemical Synthesis of Resveratrol-Piperazine Cocrystals.

The 1:1 resveratrol-piperazine cocrystal was successfully synthesized and scaled-up to 300 g scale with the mechanochemical method, as a result of investigating key process parameters, namely the solvent and the grinding time. The use of water, ethanol or ethanol-water mixtures and reaction times up to 50 min were evaluated relative to the dry grinding process. Cocrystal formation and purity were monitored through X-ray diffraction and calorimetry measurements. The dry grinding resulted in an incomplete cocrystal formation, while the use of water or water-ethanol mixture yielded a monohydrate solid phase. Pure ethanol was found to be the optimal solvent for large-scale cocrystallization, as it delivered cocrystals with high crystallinity and purity after 10-30 min grinding time at the laboratory scale. Notably, a relatively fast reaction time (30-60 min) was sufficient for the completion of cocrystallization at larger scales, using a planetary ball mill and a plant reactor. Also, the obtained cocrystal increases the aqueous solubility of resveratrol by 6%-16% at pH = 6.8. Overall, this study highlights the potential of solvent-assisted mechanochemical synthesis as a promising new approach for the efficient production of pure resveratrol-piperazine cocrystals.

Methods and means of ensuring the accuracy of bending deformation measurements of connecting pipe between the VVER reactor and the steam generator in the conditions of a fuel campaign

The welded joint zone (WJ) No. 111 is a welding unit for the "hot" collector of the primary coolant to the DN1200 steam generator nozzle. The experience of operation of steam generators PGV-1000M of NPP power units with VVER-1000 shows that the WJ zone No. 111 is prone to the formation and development of operational defects. According to the results of periodic non-destructive testing of the WJ No. 111 zone performed at the PG of various power units of the VVER-1000 NPP, cases of detection of extended planar-type discontinuities, including through ones, were repeatedly recorded. Thus, this zone is one of the most critical zones of the reactor plant (RU) with VVER -1000. The root cause of the formation and development of operational defects has not yet been identified. Solving the problem of cracking of WJ No. 111 steam generators of nuclear power plants with VVER-1000 is one of the priority areas for improving the safety of operation of the power unit during the over-design service life. This problem is determined by a complex combination of temperature conditions, mechanical and corrosive effects, is relevant and has not yet been definitively solved. Therefore, it is very important to ensure the collection of reliable data on the actual stress-strain state of the WJ No. 111 zone in various operating modes (pressure loading of steam generators, heating/cooling of the power unit, hydraulic testing of the 1st and 2nd circuits). The article reveals the approaches and stages of creating and applying a strain gauge installation to ensure metrological suitability in the conditions of a campaign fuel company at an NPP power unit by using an adequate physical model to obtain additive and multiplicative correction coefficients.

LIBS as diagnostics of analytical chemistry for surface mapping of complex mixed samples simulating debris inside the TEPCO's Fukushima Daiichi Nuclear Power Plant (FDNPS) reactor cores

The Nuclear Energy Agency (NEA) launched the Nuclear Education, Skills and Technology (NEST) Framework to pursue careers in the nuclear field, by exposing researcher working on these topics to international challenging project of real-world issue and by transferring the knowledge and expertise accumulated in the current generation to them through hands-on training.In this framework the 2022 edition of the NEST project offered a training educational period at the Collaborative Laboratories for Advanced Decommissioning (CLADS) located at Tomioka, Futaba District, in the Fukushima prefecture (Japan).Among the research sectors active at CLADS there is the application and development of the LIBS technique as diagnostics of analytical chemistry aiming at characterizing the debris inside the Tokyo Electric Power Company (TEPCO)’s FDNPS reactor cores after the tsunami of March 2011, which destroyed three of the six reactors of the plant. These debris need to be chemically characterized with techniques suitable to be implemented in compact, remote and radiation resistant devices, due to the residual radioactivity of the cores. The present study shows the results in realizing chemical bi-dimensional maps of samples in the form of compressed tablets of mixed oxides, with a complex distribution and concentration of chemicals simulating these debris. The results allowed to create detailed maps of the samples, with spatial resolution down to 0.5 mm and an excellent correspondence between the real spatial distribution of the materials and that reconstructed by LIBS.Moreover, it was found a good correspondence between the nominal concentration of the chemicals and the concentration estimated by using LIBS.This study shows the potentialities of LIBS in the realization of chemical maps on samples of interest for FDNPS, providing multi-elemental images of the samples under analysis.

Systematic solution for multi-model control approaches in nonlinear dynamic systems based on the s-gap metric

This paper proposes a systematic approach for optimizing the distribution of local models in multi-model control systems (MMCS) to enhance overall robustness. While existing literature discusses this method for linear parameter varying (LPV) and uncertain linear time-invariant (LTI) systems, significant limitations persist in addressing nonlinear dynamic systems. Robust control tools like the gap metric and generalized stability margin (GSM) have limited effectiveness in analyzing the robustness of nonlinear feedback systems. To address these challenges, novel concepts of the gap metric and GSM are introduced to determine central operating points (COPs) within local operating areas (LOAs) across the total operating area (TOA). These COPs guide the extraction of affine disturbance local models (ADLMs). Additionally, an optimization problem based on the s-gap metric and GSM is presented to optimize COPs placement and LOAs boundaries. Challenges such as non-monotonic behavior of the cost function and complexity arising from the s-gap metric formulation necessitate novel solution methods. To address these, constraints are applied to the cost function, and a novel discrete optimization approach is introduced. Finally, theoretical findings are applied to the Duffing system, pH neutralization process, and continuous stirred tank reactor (CSTR) plant to evaluate the proposed method's effectiveness. This comprehensive validation across different systems underscores the versatility and practical utility of the proposed approach.