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In this paper, the performance of aluminum-based nanofluids with a possible application in pressurized water reactors is numerically investigated. A 605 mm long 4-rod array square (2 × 2) subchannel geometry with a uniform heat flux of 50 kW/m2 has been used in CFD simulation. This analysis has been carried out using the RNG k-epsilon turbulence model with standard wall function in ANSYS FLUENT 2022R1. The impact of various flow conditions and nanofluid concentrations has been examined. The effects of variable velocities on nanofluid performance have been studied using different Reynolds numbers of 20,000, 40,000, 60,000, and 80,000. The analysis was conducted with Al2O3/water nanofluid concentrations of 1%, 2%, 3%, and 4%. A comparison of the Nusselt number based on five different correlations was conducted, and deviations from each correlation were then presented. The homogeneous single-phase mixer approach has been adopted to model nanofluid characteristics. The result shows a gradual enhancement in the heat transfer coefficient with increasing volume concentrations and Reynolds numbers. A maximum heat transfer coefficient has been calculated for nanofluid at maximum volume concentrations (ϕ = 4%) and highest velocities (Re = 80,000). Compared to the base fluid, heat transfer was enhanced by a factor of 1.09 using 4% Al2O3. The Nusselt number was calculated with a minimal error of 3.62% when compared to the Presser correlation and the maximum deviation has been found from the Dittus–Boelter correlation (13.77%). Overall, the findings suggest that aluminum-based nanofluids could offer enhanced heat transfer capabilities in pressurized water reactors. [ABSTRACT FROM AUTHOR]

[Introduction] Under the scouring impact of secondary side fluid, the steam generator tube is prone to flow-induced vibration. One of the main mechanisms leading to tube vibration is the random turbulent force. When the fluctuating pressure frequency of the fluid is close to the natural frequency of the tube, structural resonance will be caused and long-term vibration will lead to the failure of the tube. Therefore, it is necessary to study the dynamic response characteristics of steam generator tubes under fluid excitation. [Method] In this paper, a flow-induced vibration test device for the bend zone of tube bundle with a pitch-diameter ratio of 1.47 was designed. The secondary side fluid condition was simulated by air-water two-phase flow. The fluid fluctuating pressure and the vibration acceleration of tube bundle with a void fraction of 0.7～0.98 and a flow velocity between tubes of 5～13 m/s was measured. [Result] The results show that: the main frequency of fluctuating pressure is close to the natural frequency of tube at low flow velocity, which is easy to causes resonance; when resonance occurs, the amplitude of the tube bundle increases; with the increase of velocity between tubes, the fluctuating pressure on tube bundle increases correspondingly. With the increase of the void fraction, the fluctuation pressure first increases and then decreases. When calculating the main frequency of fluctuating pressure under the condition of air-water two-phase flow, the coefficient in the empirical formula can be adjusted appropriately. [Conclusion] This test simulated the working conditions of the secondary side fluid operation in the bend zone of steam generator tube bundle and considered the geometric similarity with the prototype and the similarity of support and constraint in the design of the model. This test was closer to the actual situation than the previous experimental research and can provide design references for engineering applications.

This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors.

This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors. [ABSTRACT FROM AUTHOR]

Increasing the number of pins within a pressurized water reactor (PWR) assembly reduces pin temperature for a given assembly power. In conjunction with a core retrofit, this presents a potential route to PWR uprate, which is of growing interest given recent increases in electricity prices. However, most PWRs utilize regular lattice designs with fixed guide tube positions, such as the very common 17 × 17 lattice design with 25 guide/instrumentation tubes. These tubes are aligned with penetrations in the reactor pressure vessel, which presents a prohibitive obstacle to retrofit, and more widely, may "lock" many PWRs to this particular fuel configuration. In this paper, an irregular PWR fuel assembly is proposed. It is shown that a backward-compatible lattice with 324 fuel pins per assembly (BL324), uniform enrichment, and the same hydrogen-to–heavy metal ratio as a reference 17 × 17 assembly with 264 fuel pins can achieve within-assembly power peaking within 2% of the reference assembly under equivalent conditions while fixing the guide tube positions. Power peaking can be further reduced to reach that of the existing fuel assembly by reducing the enrichment of 36 of the pins by 0.2 wt%. The fuel assembly could potentially either support a significant uprate of up to ~20% in conjunction with low-enriched uranium plus (LEU+) fuel or a more aggressive cycle design, and hence, improved discharge burnup at the same power and batch strategy. A subchannel analysis shows that the coolant heat-up distribution is comparable to the reference assembly. However, the pressure drop is estimated to be 4% higher, which would challenge the performance of transition cores containing both 17 × 17s and BL324s. Further incremental changes to BL324 may be attractive, either to improve manufacturability or to slightly improve performance through formal optimization. [ABSTRACT FROM AUTHOR]

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14C has become the nuclide that contributes the most to the annual effective dose to the surrounding public among the radioactive effluents during normal operation of nuclear power plants, but for a long time, domestic nuclear power units have not been equipped with airborne 14C treatment facilities, it is important and urgent to seek technical methods to reduce 14C emissions based on potential and actual needs under the requirements of the existing emission limits. Domestic nuclear power units under construction and in operation are mainly pressurized water reactors (more than 95% of the total), and the airborne 14C in this type of reactors mainly exists in the form of alkane compounds. Therefore, in this study, 14CH4, which accounts for the largest proportion of 14C alkane compounds and has the most stable chemical properties, was taken as the treatment target, and low-temperature plasma technology was introduced to investigate its discharge behavior and CH4 treatment performance. The results show that under the optimal conditions of normal temperature and pressure, output voltage of 17.89 kV, and gas flow rate of 0.83 cm/s, the plasma’s CH4 treatment efficiency can reach 99.37%, and the CO2 selectivity can reach 46.99%. The plasma’s CH4 treatment performance can be improved by increasing the output voltage, reaction temperature, and decreasing the gas flow rate. However, since increasing the reaction temperature would bring problems about energy consumption, safety, and equipment complexity, increasing the reaction temperature is not the first choice. In addition to CO2, there are more than thirty kinds of by-products produced in the process of plasma treatment of CH4, which are dominated by organic substances. The kinetics process of plasma treatment of CH4 is in accordance with the quasi-primary reaction kinetics model, and the corresponding rate constants are 1.104 8 m3/(kW·h). The above results indicate that plasma technology has a broad development prospect in the field of airborne 14C treatment and monitoring, especially in the treatment of 14C alkane compounds. The focus of subsequent research should be focused on optimizing the reaction pathway, lowering the reaction barriers, further increasing the proportion of CH4 directed oxidation to CO2, and significantly reducing the formation of by-products.

AbstractThis research presents the development of a one-dimensional analytical model to investigate the impact of pressure variations in the primary loop on natural circulation (NC). The model takes into account a sinusoidal input heat distribution and derives equations for the parameters of NC. The model covers a broad spectrum of NC patterns, spanning from fully single-phase to fully two-phase flow. The research demonstrates a smooth and continuous transition between various kinds of NC. Moreover, the research demonstrates that NC is capable of efficiently dissipating the decay heat generated inside the core of a pressurized water reactor, encompassing a range from 100% to 60% of the total inventory present within the primary loop. The findings of this study are compared to prior research outcomes and demonstrate a reasonable level of consistency. Additionally, comparisons are made with uniform input power distribution to demonstrate that there are no significant differences in the NC parameters when using sinusoidal heat input. [ABSTRACT FROM AUTHOR]

AbstractAPOLLO3® is a French deterministic reactor code for lattice and core calculations. It has been developed at CEA (Commissariat à l’Energie Atomique) with the financial and technical support of EDF (Electricité de France) and Framatome since 2007. The main goal of APOLLO3 is to provide a unified deterministic tool that includes the capabilities of the French lattice and core codes of the previous generation, APOLLO2, CRONOS2, ECCO, and ERANOS, introducing many improvements to reactor physical modeling in a modern and flexible software architecture platform for research and development and industrial activities. This paper presents an overview of the main new capabilities of the APOLLO3 code in the lattice and the core components. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

In this study, dissimilar metal welding of A508/IN-182/IN-52M/308L/316L was employed in the fabrication of pipe nozzles for pressure vessels in pressurized water reactors (PWR). This process involved the fusion of lowalloy A508 and stainless steel 316L using gas tungsten arc welding (GTAW) welding, with the use of nickelbased alloy Inconel 182 as a filler material. Additionally, Inconel 52M served as an overlay layer, and material 308 acted as a buffer layer to facilitate bonding between 316L and the overlay welding layer. The primary objective of this study was to investigate the electrochemical aspects of galvanic corrosion and analyze the surface properties of dissimilar welds in a simulated PWR environment, specifically in a simulated hot water (SHW), over a three-month period. Potentiodynamic polarization tests were conducted on the testing solution, which contained a mixture of inhibitors, including sodium molybdate (Na2MoO4) and sodium nitrite (NaNO2). The surface morphology was thoroughly examined using optical microscopy and scanning electron microscopy - energy dispersive spectroscopy (SEM-EDS), while phase analysis was performed using x-ray diffraction (XRD) testing. The results indicated that the most efficient inhibitor combination treatment for dissimilar metal welds achieved an 89.98 % reduction in corrosion on A508 metal. This was accomplished by utilizing a mixed concentration of 6000 ppm sodium molybdates combined with 4000 ppm sodium nitrites. As a result, the corrosion rate of A508 decreased significantly from 0.475 mpy to 0.047 mpy, bringing it in line with the corrosion potential of other alloys and establishing a passive zone. In summary, the corrosion rate of A508/IN-182/IN-52M/ 308L/316L in PWR simulations decreased as the concentration of corrosion inhibitor, specifically sodium molybdate and sodium nitrite, increased. When the inhibitor dosage concentration was optimized, a protective thin molybdenum-containing film formed on the surface, effectively guarding against random pitting caused by galvanic corrosion. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

321 stainless steel (SS) is one kind of austenitic stainless steel, in which Ti is added as stabilizing element. Due to its excellent corrosion resistance and comprehensive mechanical properties under high temperature, 321SS is an important construction material in light water reactor (LWR). It is prone to cold work due to various factors in the whole cycle, which changes its working performance, especially corrosion resistance and stress corrosion cracking (SCC) property, while corrosion and SCC are main failure forms of 321SS in LWR. The mainstream view is that the transformation of grain boundary type, the increase of slip lines, the formation of voids and the phase change from austenite to martensite caused by cold work will increase the probability of failure. However, the dominant factors and main mechanisms are still controversial. The purpose of this study is to investigate the effect of cold work on corrosion and SCC of 321SS and preliminarily explain the mechanism. Samples with different work rate and martensite fraction were acquired through different extent of cold work and hot work. The as-received specimen was remarked as SA, the cold worked specimen was remarked as CW20 and the hot worked HW20. The corrosion behavior was studied by high temperature electrochemical tests. The electrochemical impedance spectroscopy (EIS) of SA and CW20 was measured in a simulated pressurized water reactor (PWR) first circuit environment, and HW20 was used as a comparison. The SCC performance of specimens was tested by stress corrosion test under slow strain rate tensile (SSRT) loading mode. After tensile tests, micro cracks initiated on the surface were counted and the average crack length was calculated to evaluate SCC sensitivity. The microscopic characteristics of the samples were analyzed by X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive spectrometer (EDS). XRD analysis shows that barely no martensite exists in SA, the martensite volume fraction in CW20 is 22% and 11% in HW20. In EIS tests, the charge transfer resistance rises both in CW20 and HW20 compared with SA. SA has the highest film resistance, HW20 the second, and CW20 has the lowest. After SCC test, the most cracks initiated on the surface of HW, and the least on SA. The conclusions can be drawn as following: The cold work causes the transformation from austenite to martensite in the matrix, and the high temperature inhibits this process. With the increase of strain degree, the charge transfer resistance increases, and the film resistance decreases with the increase of martensite content. A Cr-depleted zone is caused by martensite at the interface between the oxidize layer and the matrix, thus leads to the reduction in protective effect of the passivation film. With the increase of martensite content, the film resistance decreases, and the corrosion resistance of the passivation film decreases. Micro cracks tend to initiate more easily after either cold work or hot work compared with primary solution annealed 321SS. The martensite phase distributed in the matrix is oxidized preferentially and suppresses the corrosion of austenitic phase. Under the conditions of this research, strain-induced martensite inhibits SCC crack initiation.

BackgroundIn thermal pipelines of nuclear power systems, thermal stratification is a common phenomenon that can cause stress concentration and deformation of pipeline structures, thereby leading to safety hazards. A stagnant branch pipe is connected to the main coolant pipe, and a large temperature difference exists between the fluid in the pipe and the coolant in the main pipe of the primary circuit. Due to factors such as turbulent flow penetration and valve leakage, thermal stratification is prone to occur in the branch pipe.PurposeThis study aims to analyze the temperature change characteristics and flow characteristics of thermal stratification in stagnant branch pipes and provide a theoretical basis for subsequent experimental research and stress analysis.MethodsFirstly, a stagnant branch pipe model was established, and numerical simulation of thermal stratification phenomenon in stagnant branch pipes was conducted using FLUENT 2022 to analyze the temperature variation characteristics of the pipe wall and the distribution characteristics of the flow field inside the pipe. Then, the SST k-ω model was used to perform three-dimensional numerical simulation of the thermal stratification of stagnant branch pipes, with a leakage flow rate of 0.062 kg·s-1, leakage temperature of 488.15 K, and leakage pressure of 6 MPa.ResultsThermal stratification is prone to occur in horizontal pipe sections. Without insulation measures and a large pipe diameter, thermal stratification can be exacerbated, while the curved section can effectively reduce the temperature difference of the cross-section. A backflow phenomenon occurs in the horizontal section of the stagnant branch pipe, while the structure of the large and small end pipe sections causes secondary backflow in the flow field inside the pipe. The backflow phenomenon is not conducive to the mixing of cold and hot fluids in the pipe; consequently, the influence time of thermal stratification is longer.ConclusionsA significant difference in the thermal stratification phenomenon exists between the stagnant branch pipe and equal cross-section pipes.

In this paper, energy, exergy, economic analysis and multiobjective optimization of a pressurized water reactor (PWR) nuclear cogeneration plant for hydrogen production through high-temperature steam electrolysis (HTSE) are carried out. HTSE requires energy in the form of both heat and electrical work. A novel parameter, namely the heat/total energy ratio, is defined, and used as a decision variable in optimization. In addition to energy ratio, hydrogen production capacity, reactor thermal power, live steam temperature, reheating mass flow rate ratio, reheating temperature and steam extraction location are considered as the decision variables to simultaneously optimize the thermal efficiency, thermal-to-hydrogen efficiency, utilization factor, exergy efficiency and total revenue of the cogeneration plant. The analysis and optimization focus on the secondary cycle of the PWR and the effects of hydrogen and electricity prices and ambient conditions are also taken into account since these prices have a significant impact on the optimum design. For a hydrogen price of 4 $/kg and an electricity price of 0.1 $/kWh, when equal preference is given to all objective functions, the optimum production capacity is 6.778 kg/s. The energy ratio has an optimum value if the optimization focuses exclusively on the thermal efficiency and total revenue. • Thermodynamic analysis and multiobjective optimization of a PWR nuclear hydrogen production plant are performed. • A novel parameter, heat/total energy ratio is defined for the high-temperature steam electrolysis process. • Increase in energy ratio improves thermal-to-hydrogen efficiency, exergy efficiency, utilization factor and total revenue. • For a hydrogen price of 4 $/kg and an electricity price of 0.1 $/kWh, the optimum production capacity is 6.778 kg/s. • Price independent optimization yields an optimum hydrogen production capacity range of 3.826–5.707 kg/s. [ABSTRACT FROM AUTHOR]

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation Exercise on past fuel melting irradiation experiments, organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Framework for IrraDiation ExperimentS (FIDES) framework by the Core Group (CEA, EDF, and SCK‧CEN) and open to all FIDES members. The exercise consisted in simulating two past power ramps where fuel melting was detected: (1) the xM3 staircase power transient [ramp terminal level (RTL) 70 kW‧m−1, average burnup 27 GWd‧tU−1], carried out in 2005 in the R2 reactor at Studsvik (Sweden), where the rodlet maintained its integrity, and (2) the HBC4 fast power transient (RTL 66 kW‧m−1, average burnup 48 GWd‧tU−1), carried out in 1987 in the BR2 reactor at SCK‧CEN (Belgium), where the cladding failed during the experiment. The exercise was joined by 13 organizations from 9 countries using 11 different fuel performance codes. In this paper, the main results of the Simulation Exercise are presented and compared to available postirradiation examinations (PIE) or on-line measurements during the power ramps (fuel and clad diameters, rod elongation, pellet-clad gap, and fission gas release). Since the focus of the Simulation Exercise is on fuel melting assessment, determination of the boundary between melted/nonmelted fuel and the consequent definition of a melting radius from PIE are first discussed. During the HBC4 ramp, fuel melting was predicted by most of the codes despite differences in the melting models. Higher discrepancies were observed for the xM3 rod that can be attributed partly to power uncertainty and partly to the limited capability of the models to describe partial melting of the fuel during this ramp. Finally, possible code developments to improve simulation results are presented. [ABSTRACT FROM AUTHOR]

This paper presents 2D(r, $\it \theta $ θ) simulations of the HBC-4 power-to-melt experiment performed with the fuel performance code ALCYONE. The HBC-4 experiment is one of the two test cases selected for the simulation exercise on past fuel melting experiments of the Power to Melt and Maneuverability (P2M) project. The ramp terminal level (RTL) at peak power node (PPN) has been estimated at 66 kW·m−1 by gamma scanning and 70 kW·m−1 based on online measurements of thermal fluxes. The fuel burnup at PPN was close to 60 GWd/tU−1. The cladding failed during the short holding time at a RTL of 40 s. Fuel melting took place at the pellet center, and in particular, in front of clad cracks. In this paper, simulations of the HBC-4 power-to-melt experiment are performed using an updated version of the 2D(r, $\it \theta $ θ) scheme of ALCYONE where half of the fuel pellet is described. This configuration allows for the modeling of clad failure by iodine stress corrosion cracking and of its consequences on fuel pellet deformation. The modeling of fuel melting relies on thermochemical equilibrium calculations performed with the OpenCalphad Gibbs Energy Minimizer and the Thermodynamics of Advanced Fuels International Database. The simulation without clad failure indicates that the solidus is reached during the HBC-4 experiment but not the liquidus. The simulation with clad failure leads to a small increase in the fuel temperature that is sufficient to reach the liquidus at the pellet center, in agreement with postirradiation examination (PIE). The impact of water ingress in the rod and vaporization at the pellet surface is discussed, showing that it could explain the pronounced swelling of the fuel pellet reported from the PIE. [ABSTRACT FROM AUTHOR]

Radionuclides emitted by nuclear power plants may have effects on the environment and public health. At present, research on radioactive material effluent in the industry mainly focuses on the treatment of radioactive effluent and the particle size distribution of the primary circuit. There is little research on the particle size of radioactive material during the migration process outside the primary circuit system, as well as the flocculation precipitation and other enrichment phenomena during the collection process of effluent. Therefore, this study relies on the sampling of effluent from an in-service nuclear power plant to measure its radioactivity level by particle size range. At the same time, the mixing process of effluent is simulated in the laboratory to simulate the adsorption behavior of effluent during the migration process. It was found that in the activity concentration of detectable radioactive nuclides in the effluent samples, more than 95% of radioactive nuclides exist in the liquid with particle sizes less than 0.1^m, while particle sizes greater than 0.45 ^m account for less than 5%. After the sample was filtered by the demineralizer, the radioactive activity decreased. The flocculation precipitation in the waste liquid of the waste water recovery system has a certain contribution to the enrichment of nuclides. With the extension of time, the enrichment of transition elements such as cobalt and manganese is particularly obvious, so that it is distributed in the liquid again with a large particle size. In addition, large particle size substances such as colloids in seawater have a certain adsorption effect on radionuclides, which will lead to its aggregation effect again. [ABSTRACT FROM AUTHOR]

Alpha particle-emitting radiopharmaceuticals are in high demand for use in targeted alpha therapy. Ac-225 is currently produced using Th-229, but its annual production remains low, approximately 63 GBq. Previously, we produced a large amount of Ac-225 via the (n,2n) reaction in fast reactors; however, it required repetitive irradiation. In this work, we investigated a method to produce Th-229 via the (3n,x) reaction through long-term irradiation using neutrons from Pressurized Water Reactors. As target nuclides, Ra-226, which is commonly used for Ac-225 production, and Th-230, which is not widely used but is abundant, were selected. The evaluation was conducted under mixed conditions of Th-230 and Th-232. Ra-226 and Th-230 produce Th-229 (T1/2 = 7920 years) after long-term neutron irradiation. Th-229, which has a long half-life, the α-decays to produce Ra-225, and the β-decays of Ra-225 to produce Ac-225. These processes are semi-permanent owing to the long half-life of Th-229. Further, an irradiation method that does not require major changes in the upper part of the PWR fuel assembly geometry was employed by replacing the plugging device attached to the control rod guide tube with a target pin. The PWR loaded with abundant natural thorium target and irradiated with thermal neutrons for as long as approximately 5 years can produce more than twice the current world supply of Ac-225 annually and permanently. [ABSTRACT FROM AUTHOR]

With the increase of fuel cycle period of the new generation pressurized water reactor, the influence of corrosion product deposition in the core is becoming more and more serious. The cladding surface temperature and species concentration changes caused by deposits (also called Chalk River Unidentified Deposit (CRUD)) increase the risks of core power shift induced by boron accumulation, radiation dose increase and cladding corrosion. In order to predict the risks induced by the CRUD on the cladding in the pressurized water reactor, a multi-physics process model was established and optimized for the heat transfer, mass transfer, fluid flow and chemical process of the cladding surface with CRUD. The heat transfer parameters and mass transfer parameters on the cladding surface were carried out under different operating conditions, hydrochemical conditions and CRUD structures. Combined with the relationship of multi-physics processes, the effects of the heat flux, the coolant temperature, the boron and lithium concentrations in coolant and the structural parameters of CRUD on the performance of the cladding surface were discussed, and the essential reasons of CRUD induced boron accumulation were discussed. The risk equation to evaluate the CRUD induced boron accumulation under different operating parameters, hydrochemical parameters and CRUD structural parameters was proposed. The results show that when boiling occurs in CRUD, the temperature distribution can be obtained more accurately by considering the volatilization of boric acid. The presence of CRUD leads to a dramatic increase of the outer surface temperature of the cladding. The increase degree of cladding temperature increases with the CRUD thickness, the CRUD porosity and the boron concentration, and decreases with the increase of heat flux, coolant temperature and CRUD stack density. The thermal driving force from cladding to coolant increases when the CRUD thickness is less than 50 μm, the heat flux is 0.4-1 MW/m2 and the coolant temperature is less than 300 ℃. Beyond the above ranges, the presence of CRUD reduces the heat transfer from cladding to coolant. Under the classical pressurized water reactor condition, the boron accumulation in the void structure of 40 μm CRUD mainly derives from boric acid (0.057 4 g/m2) and the adsorbed boron (4.61×10-3 g/m2). The deposited boron in the void structure of CRUD in the form of Li2B4O7 is 8.34×10-5 g/m2, while LiBO2 can not deposit in 40 μm CRUD. At 1 MW/m2, a power shift of 1.5% occurs when 1/8 of the core attaching to 80 μm thick cladding CRUD. The effects on the CRUD induced boron accumulation risk from large to small are as follows: CRUD thickness, coolant temperature, pore diameter, coolant boron concentration, heat flux, porosity, while the coolant lithium concentration and chimney density of CRUD have little effect.

Radionuclides emitted by nuclear power plants may have effects on the environment and public health. At present, research on radioactive material effluent in the industry mainly focuses on the treatment of radioactive effluent and the particle size distribution of the primary circuit. There is little research on the particle size of radioactive material during the migration process outside the primary circuit system, as well as the flocculation precipitation and other enrichment phenomena during the collection process of effluent. Therefore, this study relies on the sampling of effluent from an in-service nuclear power plant to measure its radioactivity level by particle size range. At the same time, the mixing process of effluent is simulated in the laboratory to simulate the adsorption behavior of effluent during the migration process. It was found that in the activity concentration of detectable radioactive nuclides in the effluent samples, more than 95% of radioactive nuclides exist in the liquid with particle sizes less than 0.1μm, while particle sizes greater than 0.45 μm account for less than 5%. After the sample was filtered by the demineralizer, the radioactive activity decreased. The flocculation precipitation in the waste liquid of the waste water recovery system has a certain contribution to the enrichment of nuclides. With the extension of time, the enrichment of transition elements such as cobalt and manganese is particularly obvious, so that it is distributed in the liquid again with a large particle size. In addition, large particle size substances such as colloids in seawater have a certain adsorption effect on radionuclides, which will lead to its aggregation effect again.

Heat and mass transfer and gas distribution in a nonequilibrium steam-water volume with a phase separation level containing a noncondensable gas are considered. These processes relate primarily to the steam–gas pressurizer (SGP) of the integral reactor and determine the behavior of the SGP in normal and emergency modes. Similar processes occur in a pressurized water reactor and in a containment volume at the final stage of emergency regimes with loss of coolant. For small-power reactors of the RITM-200 type with a gas pressurizer, it is necessary to take into account the phenomena associated with the behavior of noncondensable gases. A brief description is given of the GARRIC 2.2 computer program for calculating the distribution of noncondensable gases in the primary circuit and the characteristics of the SGP of a pressurized water reactor of an integral layout. The description of the processes of heat and mass transfer and gas distribution is based on the results of experiments carried out on SGP models in a wide range of parameters and gas contents. Calculation of mass transfer, including surface evaporation and steam condensation, and gas transfer on the surface is carried out using the analogy model of heat and mass transfer under conditions of natural convection of the medium. The results of the verification of the GARRIC 2.2 program on the experimental data obtained in the study of vapor condensation on the walls, evaporation on the free surface of water, and gas distribution in the steam–gas volume at full-scale parameters are presented. The GARRIC 2.2 program was certified by Rostekhnadzor in 2014. [ABSTRACT FROM AUTHOR]

The extension of the operating domain of PreussenElektra's Konvoi-type pressurized water reactors (PWRs) beyond the natural end of cycle is known as stretch-out operation. A range of possibilities exists to increase nuclear fuel utilization to continue operation after the boron concentration reaches its dilution limit. The most basic option is to continue operation with constant average moderator temperature, which results in a relatively fast decrease in reactor power. From a fuel utilization point of view, this is the least optimal procedure. In PreussenElektra's PWR fleet, an enhanced operation mode is adopted, leading to a comparatively modest decrease in reactor power and very high utilization of nuclear fuel. Initially, the stretch-out mode provided an option to gain flexibility regarding outage planning. More recently, the stretch-out method has served as a practical approach to optimizing electricity generation costs during the last cycles before the final shutdown as stipulated by law, as operators can extend the cycle length in a range of 30 to 60 days after the natural end of cycle. This paper describes the licensing rationale, the feasibility of this type of operation, and the operating requirements and experience. The system parameters affected by stretch-out operation are discussed. Adjustments of set points of thermal-hydraulic variables in the primary and secondary systems are explained. Licensing requirements for safe reactor operation in stretch-out mode are reviewed. Furthermore, aspects of neutronic and thermal-hydraulic core surveillance are included. After more than 35 years and counting, the methods of increasing fuel utilization are not new, and an evaluation of experience and effectiveness is in order. [ABSTRACT FROM AUTHOR]

The replacement of LiOH with KOH for pH control in the primary water of pressurized water reactors (PWRs) is being considered due to the growing cost of enriched 7Li. This study aims to investigate the susceptibility of the primary water stress corrosion cracking (PWSCC) initiation in simulated primary water with KOH. Firstly, the susceptibilities in KOH and LiOH chemistries were compared by conducting uniaxial constant load tests on Alloy X-750 at 360°C. The results showed no significant difference in the time to initiation between the two chemistries. Secondly, the effect of dissolved hydrogen (DH) concentration in the KOH chemistry was examined. The DH concentration of the test water was varied at 5, 30 or 45 ml/kg-H2O. The results showed that the time to initiation of PWSCC was significantly extended under the low DH condition (5 ml/kg-H2O). This observed effect of DH concentration in the KOH environment agrees with the previously reported effect observed in the conventional LiOH environment. To investigate the mechanisms underlying the PWSCC tests, the oxide films on the test specimens were characterized using electron microscopes. The oxide films formed in the KOH and LiOH chemistries did not show significant differences. Additionally, under the low DH condition, the occurrence of selective internal oxidation under the inner oxide film was relatively minor. The results suggest that the use of KOH would not have an adverse effect on PWSCC initiation, and PWSCC initiation can be mitigated with a low concentration of DH in the KOH environment as well as in the LiOH environment.

Fission gas isotopic compositions are sensitive to a variety of reactor operating parameters that include the neutron flux, neutron fluence at different neutron energies, and operating temperature. Measurements of fission gas isotopic compositions thus have potential to constrain reactor simulations for nuclear forensics and safeguards applications. In this paper, we present Kr and Xe isotope measurements from a suite of samples obtained from locations that span the axial length a fuel pin with a well-characterized irradiation history and compare these data to spatially resolved reactor simulations. We observed positive correlations between fluence sensitive isotopic ratios and burnup and between a flux sensitive ratio and power, although some discrepancies are observed between the measured data and model predictions. These differences may be due to simplifications in the model and/or inaccuracies in the cross sections. A much broader measurement to model comparison is required to better understand the discrepancies. [ABSTRACT FROM AUTHOR]

The salt hideout phenomenon of boiler water attracted the close attention of specialists as long ago as the 1940s–1950s. By the end of the 1980s, the majority of researches had arrived at the conclusion that the governing role in the hideout phenomenon is played by the deposits of structural material corrosion products (crud) on the steam-generating surfaces of the equipment of nuclear and thermal power plants. The steam-generation process takes place under confined conditions, which causes degraded mass transfer between the flow core and the heat-transfer surface. This results in that water impurities concentrate in the pores of deposits and even precipitate in a solid phase form. As the steam boiler/steam generator power output increases, the concentrations of certain impurities and chemical agents in boiler water decrease; this effect is called hideout, and as the load decreases, their concentrations increase (hideout return). In the last decades, a few physical and mathematical models have been developed in which the hideout phenomenon is considered from the viewpoint of boiler water impurities becoming concentrated not in the layer of permeable deposits but in the viscous sublayer of liquid at the steam-generating surface. Thus, the thermodynamic model rests on the postulates of nonequilibrium thermodynamics and is descriptive in nature. The mass-transfer model based on the laws of mass and energy conservation in the viscous sublayer incorporates an analytical expression for the impurity concentration ratio. However, this model also in fact contains only a qualitative description of the hideout process without performing its detailed comparison with experimental data. The article presents an analysis of these models and their comparison with reliable data obtained by domestic and foreign researchers, and it is shown that the key statements laid down at the essence of models based on impurity concentration in the liquid viscous sublayer are erroneous in nature. Adequate fundamental principles of mass transfer under hideout conditions are of significant theoretical and practical importance for working out operation regulations and securing reliable operation of installations with boiling coolant at nuclear and thermal power plants. [ABSTRACT FROM AUTHOR]

During the normal operation of pressurized water reactors, tritium has contributed more than 95 % of the total radioactivity of all the radionuclides discharged into the environment and has been recognized as the key radionuclide in the design and operation of reactors. In the paper, the tritium production of the AP1000 reactor by the Westinghouse approach has been introduced and the diffusion fractions of tritium from fuel rods in the approach have been reviewed according to advances in research on the diffusion of tritium from zirconium alloys. The historical tritium discharge data from the predecessors with very similar core designs of the AP1000 reactor have been collected and sorted. According to the statistical results, the average tritium discharge approximates the tritium generated from neutron activation of boric acid and lithium hydroxide and it shows that the release fraction of tritium through fuel cladding with zirconium alloy may be neglected for the best-estimated condition. The release of tritium from secondary neutron sources has been validated by a comparison of historical tritium discharge from the predecessors with and without secondary sources. In addition, it indicates that local maximum tritium emissions from the U.S. Pressurized Water Reactors are remarkably affected by batch liquid release, especially before or during the overhauls. It will help recognize the tritium generation in the primary loops and optimize the management of radioactive emissions for the utilities of the AP1000 reactor in the future.

A pressurized water reactor (PWR) is a system of several integrated components such as the core, steam generator, hot leg, cold leg, and plenums. The subsystems consist of critical parameters and malfunctions that cause potential accidents. Therefore, a PWR requires a control system for safe and stable operation over its lifetime. In this study, the state-space model of the PWR core is established and validated with published work. Then, a beetle antenna search (BAS) algorithm–optimized radial basis function (RBF) neural network proportional-integral-derivative (PID) control (BAS-RBF-PID) strategy is proposed to regulate the core power. The BAS-RBF-PID control approach computes the control input to optimize the PWR core output power to track the reference command. The integral absolute error and integral time absolute error criterion functions are used to measure the control performance. The sensitivity of the control input on the PWR output is examined through the Jacobian, and the stability is analyzed by using the Lyapunov approach and Nichols chart. The simulation results verified that the PWR core output power chased the reference command smoothly as compared with the BAS-PID and PID strategies with good performance. This confirms that the control signal optimizes the core power effectively. This study gives the benefit to apply the BAS-RBF-PID algorithm in other nuclear engineering fields for control purposes. [ABSTRACT FROM AUTHOR]

During the normal operation of pressurized water reactors, tritium has contributed more than 95 % of the total radioactivity of all the radionuclides discharged into the environment and has been recognized as the key radionuclide in the design and operation of reactors. In the paper, the tritium production of the API000 reactor by the Westinghouse approach has been introduced and the diffusion fractions of tritium from fuel rods in the approach have been reviewed according to advances in research on the diffusion of tritium from zirconium alloys. The historical tritium discharge data from the predecessors with very similar core designs of the AP1000 reactor have been collected and sorted. According to the statistical results, the average tritium discharge approximates the tritium generated from neutron activation of boric acid and lithium hydroxide and it shows that the release fraction of tritium through fuel cladding with zirconium alloy may be neglected for the best-estimated condition. The release of tritium from secondary neutron sources has been validated by a comparison of historical tritium discharge from the predecessors with and without secondary sources. In addition, it indicates that local maximum tritium emissions from the U.S. Pressurized Water Reactors are remarkably affected by batch liquid release, especially before or during the overhauls. It will help recognize the tritium generation in the primary loops and optimize the management of radioactive emissions for the utilities of the AP1000 reactor in die future. [ABSTRACT FROM AUTHOR]

Corrosion products in pressurized water reactors are challenging to study in situ, yet understanding their properties is key to improving reactor performance and radiation reduction. In this study, a hydrothermal synthesis technique was used to produce nickel ferrite (NiFe2O4) particles from goethite (α-FeOOH) and nickel nitrate hexahydrate [Ni(NO3)2 6H2O] in the presence of sodium hydroxide (NaOH). X-ray diffraction was used for phase identification, with scanning electron microscopy used for particle shape and size analysis. By varying the [Ni]:[Fe] ratio of the precursors and synthesis temperature between 100°C to 250°C, a phase diagram was developed to determine the stability field in both composition and temperature for obtaining a single-phase, nonstoichiometric nickel ferrite product. The compositional boundaries of the single-phase region of the diagram are a function of temperature, consistent with the increased solubility and reaction rates at temperatures above 125°C. The single-phase nickel ferrite encompasses [Ni]:[Fe] ratios in a very narrow range at 150°C, only 0.35 to 0.375, but widens as a function of temperature and reaches its greatest breadth at 250°C. At this temperature, a single-phase product is obtained for a range of starting compositions from 0.30 to 0.425. Outside of this window, additional nanoparticles are obtained whose identity and composition vary with both temperature and starting mixture. On the lower nickel content side of the single-phase region, the mixture contains either unreacted goethite (for temperatures below 200°C) or hematite (α-Fe2O3) at 200°C or higher. On the Ni-rich side of the single-phase region, theophrastite [β-Ni (OH)2] was obtained along with the nickel ferrite, at all temperatures studied. The single-phase window was widest at 250°C, resulting in nickel ferrites with a Ni mole fraction between 0.23 and 0.31. [ABSTRACT FROM AUTHOR]

Natural hazard-induced nuclear accidents, such as the Fukushima Daiichi Accident that occurred in Japan in 2011, have significantly increased reactor safety studies in understanding nuclear power plant (NPP) responses to external hazard events such as earthquakes and floods. Natural hazards could cause the loss of offsite power in nuclear power plants, potentially leading to a Station Blackout (SBO) accident that significantly contributes to the overall risk of nuclear power plant accidents. Despite the fact that extensive research has been conducted on the station blackout accident for nuclear power plant, further understanding of these events is needed, particularly in the context of the dynamic nature of external hazards such as external flooding. This paper estimates the progression of station blackout events for a generic pressurized water reactor (PWR) in response to external flooding events. The original RELAP5-3D model of the Westinghouse four-loop design pressurized water reactor was adopted and modified to simulate the external flood-induced station blackout accident, including the short-term and long-term station blackout scenarios. A sensitivity analysis of long-term station blackout, examining reactor operation times and analyzing key parameters over time, was also conducted in this work. The results of the analyses, especially the critical timing parameters of key event sequences, provide useful insights about the time during the external flooding event, which is important for plant operators to make timely decisions to prevent potential core damage. This paper represents significant progress toward developing an integrated risk assessment framework for further identifying and assessing the effects of the critical sources of uncertainties of nuclear power plant under external hazard-induced events.

The optimization of water chemistry in pressurized water reactor (PWR) primary circuit is one of the most effective ways to achieve both safety and economy for operating PWR nuclear power plants (NPPs). Special attention has been paid to fundamental research and engineering application of Zn-injection water chemistry (ZWC) in PWR primary circuit in recent years in China due to the rapid development of nuclear power. The present paper mainly reports the status of PWR NPPs in China and some fundamental understandings of ZWC on material corrosion. Effects of temperature (T), pH at T value, Zn concentration, Zn-injection sequence and lasting time on electrochemical corrosion behavior, oxide film characteristic and stress corrosion cracking (SCC) susceptibility of nuclear-grade austenitic alloys were investigated. A modified point defect model was used to discuss the effects of ZWC on the oxide films in high-temperature pressurized water. Some water chemistry parameters in PWR primary coolant system are proposed for mitigating corrosion and SCC of nuclear-grade austenitic alloys.

Obtaining real time quantitative condition assessment results is a focus in the field of nuclear power system PHM. The existing researches on the nuclear power system condition assessment based on the Multi-Criteria Decision-Making framework do not consider the faults perniciousness differences. This leads to the assessment results relying on the system status deviation degrees(severity), and the weighting and aggregation processes cannot effectively reflect the differences in the consequences and risks of different faults, which represent the faults harmfulness. A quantitative assessment method based on fault severity and fault harmfulness is proposed to address this issue. The novel method quantifies the fault severity based on the similarity principle while diagnosing system faults, and then analyzes the devices mainly affected by various faults to quantifies the harmfulness. Combining the system status deviation degrees(severity), the devices and parameters importance, and the differences in faults harmfulness, the novel method aggregates comprehensive system condition assessment results. Based on a widely recognized nuclear power system faults dataset, the novel method was compared with other methods. The conclusion is that the novel method considers the differences in the faults harmfulness, resulting in more reasonable assessment results and avoiding insufficient or excessive warnings. • An novel nuclear power system quantitative condition assessment method. • The weighting and aggregation consider the faults perniciousness differences. • Avoid insufficient or excessive warnings and support efficient maintenance. [ABSTRACT FROM AUTHOR]

• Higher residual stress/strain and dislocation density in alloy 600 HAZ than in base. • Thicker oxide and deeper intergranular oxidation in alloy 600 HAZ than in base. • Longer SCC length and higher cracking engagement in alloy 600 HAZ than in base. • Stress-assisted oxidation contributes to enhanced SCC growth in alloy 600 HAZ. • The impact of local mechanical parameters on SCC in the HAZ is highlighted. The impact of welding-induced mechanics on the oxidation and stress corrosion cracking (SCC) behavior of Alloy 600 in a pressurized water reactor (PWR) primary water was investigated using an Alloy 600-Alloy 152 M weldment. The microstructural analysis found a 0.05 mm-wide composition transition zone from welding. The Alloy 600 heat-affected zone (HAZ) grain size matches the Alloy 600 base (600B). Deformation hardening initially decreases and stabilizes from the fusion boundary (FB) to the 600B Comparing the microstructural characteristics of the specimen from the HAZ at about 0.5 mm from the FB with specimen 600B shows identical composition, but the former exhibits higher Vickers hardness (HV), kernel average misorientation (KAM), dislocation density, and residual stresses. Results from both short- and long-term oxidation tests indicate that the specimen from the HAZ exhibits a greater thickness of the inner oxide film, a higher number of local oxidation sites, and an increased maximum depth of intergranular (IG) oxidation compared to the 600B specimen. Multiple sets of SCC test results demonstrate that the crack growth rate (CGR) in the specimens from the HAZ is higher than that in the base, with the former showing longer IG cracks and a greater proportion of IG cracking. [ABSTRACT FROM AUTHOR]

In this study, the corrosion behavior of pre-ion-irradiated 304 SS is compared to that of un-irradiated 304 SS in simulated Pressurized Water Reactor (PWR) conditions for both early stages of corrosion and longer exposure times. Transmission electron microscopy is used to characterize the oxide scale. The study shows that although the pre-irradiated microstructure does not change the nature of the oxide and its dual-layer structure (i.e. inner/outer oxide layers) formed on 304 SS, the path of growth of the inner oxide layer is affected by the irradiated microstructure resulting in oxide protrusions into the matrix; a mechanism is proposed. • A dual oxide structure forms on 304 SS after corrosion in PWR conditions. • The dual layer shows a continuous Cr-rich inner oxide and Fe-rich oxide islands on the surface. • Pre-irradiation to 10 dpa does not alter the dual-layer oxide structure but affects the inner oxide thickness and shape. • Inner oxide protrusions into the matrix are observed below the bigger oxide particles. • Irradiation loops aligning under local stress and dislocation channels may aid oxygen diffusion. [ABSTRACT FROM AUTHOR]

This research investigates the performance of new radiografted materials as adsorbents for ionic and colloidal 110 mAg, a prominent pollutant in pressurized water reactor (PWR) wastewater. A comparative analysis was conducted to evaluate their effectiveness against commercially available resins. Due to the increasing demand for new nuclear power plants and the growing global dependence on nuclear energy, it is crucial to minimize radioactive waste production. The study was conducted on simulated inactive and radioactive waste samples with trace amounts of 110 mAg that are hard or impossible to detect by standard analytic techniques. The synthesized materials were labeled PGE (amine functional groups) and PGM (sulfhydryl and amine functional groups). Functionalization with amine and sulfhydryl groups was chosen since it is well known that these groups strongly interact with Ag in both ionic and colloidal form. The tested materials demonstrate exceptional performance as adsorbents by completely removing (100 %) both ionic (PGM) and colloidal (PGE) forms of 110 mAg. Additionally, depending on the adsorbed species excellent elution of 110 mAg was achieved (>94 % for ions and >50 % for colloids) implying possible material reusability. Manufacturing cost of such adsorbents is orders of magnitude lower per kg (10 – 50 $/kg), compared to prices of commercial resins (100 – 10000 $/kg). These results indicate that such inexpensive materials could potentially be used in an efficient and selective way to treat PWR wastewater for 110 mAg contamination while reducing the operating cost and minimizing resulting solid waste. [Display omitted] • Radiografted materials (PGM/PGE) allow complete 110 mAg removal from simulated nuclear wastewater samples. • Total adsorption of colloidal 110 mAg (100 %) is achieved, in contrast to standardly applied processing techniques (> 98 %). • Inexpensive production, robustness and potential reusability compared to commercial resins. • Reduction in solid waste production and complete mitigation of 110 mAg environmental contamination. [ABSTRACT FROM AUTHOR]

The present work aims to introduce a nonlinear control scheme that combines intelligent feedback linearization (FBL) and a model predictive control (MPC) for a pressurized water reactor (PWR). The nonlinear plant model that is considered in this study is described by the first-principles approach, and it consists of 38 state variables. First, system identification using a dynamic neural network (DNN) structure is performed to obtain a standard affine nonlinear system. The quasi-Newton algorithm is employed to find the best DNN model. Then, an FBL is formulated to address the nonlinearity of the DNN model. An MPC controller is developed based on the FBL system to improve the system performance. The designed controller is compared with a linear MPC controller that is based on state-space models to evaluate the performance of the proposed controller. The proposed approach improves the load-following operation and offers better disturbance rejection capability than the conventional MPC. In addition, numerical measures are employed to compare and analyze the performances of the two control strategies.

In a typical pressurized water reactor, neutron detectors located outside the reactor core monitor reactor power. In addition, they are also used to measure the reactivity of the control rods. A novel approach to calculate the ex-core neutron detector response in a typical pressurized water reactor using the Monte Carlo technique is presented. A detailed ex-core model of the Krško nuclear power plant was developed using the Monte Carlo neutron transport code MCNP. Due to the location of the ex-core neutron detectors, the hybrid code ADVANTG is used to generate variance reduction parameters to accelerte the convergence of the results outside the reactor core. To use ADVANTG, the fixed neutron source had to be reconstructed from the criticality core calculation. This paper presents the sensitivity analysis of the response of the ex-core detectors to the neutron data libraries used, the description of the fixed neutron source and the ADVANTG parameters. It was found that a pin-wise description of the neutron source for at least two rows of fuel assemblies at the core periphery is necessary for accurate results. Our results show the importance of a correct description of the prompt neutron spectra in the high energy region and the impact this has on the response of the ex-core detectors. The method in which the prompt neutron fission spectra for important fission nuclides are weighted by the calculated reaction rates has been shown to be the best approximation, with deviations from the reference calculation within statistical uncertainty. The effect of nuclear data libraries on the response of the ex-core detector was investigated, and the difference between the ENDF/B-VII.0 and the ENDF/B-VIII.0 nuclear data libraries was ∼11%. When the deficient evaluation of the 56Fe isotope included in the ENDF/B-VIII.0 nuclear data library was replaced by the improved evaluation from the IAEA INDEN project, the differences decreased to ∼3.7%. In addition, neutron flux redistributions due to control rod movement were investigated and flux redistribution factors were updated using Monte Carlo particle transport methods. The reaction rate redistribution factors obtained with methods presented in this paper are within 1% agreement with the currently used factors.

This paper proposes a fiber optic Fabry-Perot (F-P) strain sensing system using non-scan correlation demodulation applied to the health monitoring of the pressurized water reactor's fuel assembly structures. The structural design and sensing mechanism analysis of the sensor were carried out, and the strain transfer model from the fuel sheet to the strain gauge was established. After the sensor fabrication and installation, the static tests have been conducted, and the results indicate that the sensing system can accurately measure the microstrain with a sensitivity of up to 12.6 nm/με at a high temperature (300 °C). The dynamic testing shows that the sensing system has a good frequency adaptation at 10–500 Hz. Thermal-hydraulic experiments show that the sensing system can run stably in a nuclear reactor, with high temperature, high pressure, and high-velocity flow flushing; additionally, the consistency deviation of the measured data is less than 1.5%. [ABSTRACT FROM AUTHOR]

The aim of this work is to introduce the novel concept of an m-polar fuzzy soft set, including various types of algorithms and their fundamental operations. We created mathematical modeling to analyze operational rules and discuss the advantages, disadvantages, and natural aspects of algorithms for six types of nuclear power plants. It has been determined that emerging trends and the benefits of algorithms are increasing step by step. The suggested modeling with an m-polar fuzzy soft set is integrated into the fuzzy mean environment to analyze the effect of the correlation between decision factors and decision results without an excessive duty cycle, thus minimizing energy use and other adverse effects. Based on a new group decision-making technique considering an asymmetric weight vector, we proved that Gas Cooled, Graphite-Moderated, and Pressurized Water Reactors are the optimal choices for nuclear power plants. In the end, a numerical illustration is provided for selecting the best photo to demonstrate the use of the generated technique and to exhibit its adequacy. [ABSTRACT FROM AUTHOR]

The direct current potential drop (DCPD) technique may be used in crack propagation tests to measure the crack growth rate (CGR). Potential probes attached to the specimen allow the variation of the crack length to be estimated. In this research, the DCPD technique using one single potential probe was applied to solid bar specimens (i.e., without any initial notch or crack) subjected to low-cycle fatigue testing in a simulated pressurized water reactor (PWR) environment. This particular analysis had two associated difficulties, the first one being the fact that crack initiation sites are not known beforehand, and the second one consisting in the experimental difficulties and conditioning factors associated with the simulation of the PWR environment. Nine solid bar specimens were tested to fatigue failure under different strain amplitudes and frequencies, while also measuring the corresponding DCPD signal during the fatigue process. It was observed that the initiation of multiple cracks was detected by the DCPD measurements. Moreover, as fatigue continued, one of the cracks became dominant and progressed to cause the specimen failure. The DCPD technique allowed the average CGR of the dominant crack to be estimated. Finally, the obtained average CGRs were validated by comparing them with average CGRs derived from striation spacing measurements, obtained from scanning electron microscopy (SEM) and from literature values gathered in the NUREG/CR-6909 document. [ABSTRACT FROM AUTHOR]

Background: Environmental-assisted fatigue (EAF) is a major issue for the long-term survival of nuclear power plant fleets in the U.S. and worldwide. Multi-material welded regions (e.g., nozzles) and other high-stress regions of reactor coolant system (RCS) components are prone to EAF-related damage. Objective: The discussed work describes a system-level finite element (FE) model of RCS components of a pressurized water reactor (PWR). This is with the goal of predicting the stress hotspots, strain residuals, strain amplitudes and the resulting fatigue lives. Methods: The FE model was developed considering system-level loading conditions (under connected system thermal–mechanical boundary conditions). Thermal–mechanical stress analysis was performed considering thermal stratification and a design-basis reactor loading cycle. Based on the FE model results, the strain residuals, strain amplitudes and resulting fatigue lives of RCS components were predicted. Results: The results show that some of the RCS components can have significantly different strain amplitudes, residual strain, and fatigue lives, despite having similar geometry and material. Higher residual strain can lead to accelerated cyclic hardening of material and the associated effect of EAF. The simulated component-level strain profile (under realistic multi-axial-multi-physics loading cycle) can guide the selection of appropriate test inputs for conducting laboratory-scale EAF tests, which is a focus of future works. Conclusions: Despite similar geometry and material the RCS component can have significantly different strain profiles and resulting fatigue lives. The discussed approach can help to identify and prioritize the RCS components for conducting expensive nondestructive evaluation (NDE) inspections. [ABSTRACT FROM AUTHOR]