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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]

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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.

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.

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.

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.

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.

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]

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.

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]

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]

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were

A study was performed comparing the performance of UO2 and (Th,U-233)O2 fuel in small long-life PWR. The neutronic calculation carried out by PIJ module of SRAC2006 was done to a fuel cell in 10 years of operation. The calculation was conducted by varying the enrichment of U-235 in UO2 and U-233 in (Th,U-233)O2 for 1% - 20% and also by varying the fuel volume fraction for 40%, 45%, 50%, 55%, and 60%. The performance was observed by comparing the enrichment needed by each fuel type to gain criticality in 10 years, the infinite multiplication factor (k-inf) value, and the conversion ratio (CR) value. The calculation results showed that 60% fuel volume fraction gave critical conditions with the lowest infinite multiplication factor and highest conversion ratio for both fuel types. While in terms of fissile nuclide enrichment needed, (Th,U-233)O2 had better performance than UO2, because only 5% U-233 was needed in (Th,U-233)O2 while UO2 needed 9% U-235 to gain criticality in 10 years of operation.

BackgroundSilicon carbide (SiC) composite claddings are candidate solutions for accident resistant fuel claddings in light water reactors.PurposeThis study aims to estimate the failure probability of a double-layer structured SiC cladding under a loss-of-coolant accident (LOCA).MethodsBased on a failure probability calculation method for SiC composite cladding, a quasi-steady state method was used to simulate and calculate the SiC composite cladding failure probability under transient conditions. Sensitivity analysis of the two characteristic parameters of Weibull distribution was performed by analyzing the proportion of various stresses under accident conditions. The effects of different burn-up conditions on the failure probability were investigated, and the failure probability of the cladding under different layer thickness ratios was simulated.Results & ConclusionsSimulation results indicate that the transient failure probability of SiC composite claddings is significantly affected by changes in the proportion of the composite layer and Weibull parameter, as well as the occurrence of LOCAs under different burn-up conditions. This study makes contribution to the development and design of accident resistant fuel claddings, providing reference for further investigations on the failure probability of SiC composite claddings.

Fouling due to magnetite deposition has been a major concern for steam generator (SG) tubing of pressurized water reactors (PWRs). Alloy 690 SG tubes are now used for new plants or are scheduled to replace old Alloy 600 tubes of operating plants. The purpose of this study is to investigate the magnetite deposition behavior on the two different SG tube materials: Alloy 600 and Alloy 690. Deposition tests were conducted under a sub-cooled nucleate flow boiling condition in simulated secondary water of a PWR at 270 °C. After these tests, we observed that the tube surfaces were covered with deposits composed of porous magnetite particles. We found approximately 30% more magnetite deposits on Alloy 600 than on Alloy 690. The electrostatic repulsive force between the magnetite particles and the Alloy 600 surface was only half of that between the magnetite particles and the Alloy 690 surface, resulting in an increase in the deposit mass. [ABSTRACT FROM AUTHOR]

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency.Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

The interaction mechanisms between polyacrylic acid (PAA) and Alloy 800 were investigated. Alloy 800 formed a bilayer oxide structure in all conditions, with an outer layer of Fe-rich spinel and α-Ni(OH) 2 and a Cr-rich inner layer. The chelating effect of PAA on Fe ions reduced the presence of Fe-rich spinel in the outer layer and increased the Cr content in the inner layer. However, thermal degradation of PAA (≥ 500 ppb) promoted dissolution of Cr in the initial inner layer, increasing the initial corrosion rate. The optimal PAA dosage in steam generators is determined to be approximately 250 ppb. • The formation of the bilayer oxide structure on Alloy 800 is not affected by the presence of PAA. • PAA promotes the depletion of Fe in the outer layer and the enrichment of Cr in the inner layer. • Excess PAA (≥ 500 ppb) accelerates the initial corrosion rate due to thermal decomposition. • PAA at around 250 ppb is the optimal dosage in SG as it does not adversely impact corrosion behavior. [ABSTRACT FROM AUTHOR]

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

The microstructure and texture of three 316L foils of 25 μm thickness, which were subjected to different manufacturing process, were systematically characterized using advance analytical techniques. Then, the electrochemical property of the 316L foils in simulated pressurized water reactor (PWR) solution was analyzed using potentiodynamic polarization. The results showed that final rolling strain and annealing temperature had evident effect on grain size, fraction of recrystallization, grain boundary type and texture distribution. It was suggested that large final rolling strain could transfer Brass texture to Copper texture; low annealing temperature could limit the formation of preferable orientations in the rolling process to reduce anisotropy. Potentiodynamic polarization test showed that all samples exhibited good corrosion performance in the simulated primary PWR solution.

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design- basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCMfuel in a 17 X 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to- diameter ratio (P/D) and downselected two alternate assembly designs: 11 X 11 (1.65-cm outer diameter and 1.18 P/D) and 9 X 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

Applying an accurate parametric prediction model to identify abnormal or false pressurizer water levels (PWLs) is critical to the safe operation of marine pressurized water reactors (PWRs). Recently, deep-learning-based models have proved to be a powerful feature extractor to perform high-accuracy prediction. However, the effectiveness of models still suffers from two issues in PWL prediction: the correlations shifting over time between PWL and other feature parameters, and the example imbalance between fluctuation examples (minority) and stable examples (majority). To address these problems, we propose a cost-sensitive mechanism to facilitate the model to learn the feature representation of later examples and fluctuation examples. By weighting the standard mean square error loss with a cost-sensitive factor, we develop a Cost-Sensitive Long Short-Term Memory (CSLSTM) model to predict the PWL of PWRs. The overall performance of the CSLSTM is assessed by a variety of evaluation metrics with the experimental data collected from a marine PWR simulator. The comparisons with the Long Short-Term Memory (LSTM) model and the Support Vector Regression (SVR) model demonstrate the effectiveness of the CSLSTM.

BackgroundUnder the condition of sub-cooled nucleate boiling (SNB), corrosion products in primary coolant of nuclear reactor will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Previous literature shown, zinc injection in primary coolant is an important method to inhibit the fuel crud deposition on the fuel cladding surface.PurposeThis study aims to investigate the influence of zinc concentration on the behavior of fuel crud deposition, and eventually provide guidance for zinc injection in primary coolant of nuclear power plant.MethodsThe fuel crud deposition tests of domestic zirconium alloy fuel cladding in different zinc concentrations were carried out by using a self-made fuel crud deposition device. Tubular crud deposition test specimen with built-in heating unit was designed and prepared for simulation study. After the tests, stereo microscope (SM) and scanning electron microscope (SEM) were employed to observe the macro and micro morphology of fuel crud whilst the composition of of fuel crud was observed and analyzed by the energy dispersive spectroscopy (EDS) with SEM, and X-ray photoelectron spectroscopy (XPS) was used to analyze the contents of Zn and B elements in the crud phase and inside the crud.ResultsObservation results show that the chimney-like crud formed on the fuel cladding surface becomes less obvious with increasing the zinc concentration in the coolant and the crud surface becomes flatter. Simutabeously, the crud thickness, the ratio of Ni/Fe and the boron precipitation mass within the crud are decreasing with increase of the zinc concentration. When the zinc concentration increases to 100 μg‧L-1, new Zn-containing phases precipitate within the crud.ConclusionsWithin the zinc concentration of 0~100 μg‧L-1, zinc injection in primary coolant of reactor can significantly inhibit the crud deposition on the fuel cladding surface.

BackgroundCompared with the traditional pressurized water reactor (PWR), the core design of large advanced PWR CAP1400 has significant changes, such as the increase in the number of fuel assemblies, the increase in reactor power, the increase in the average temperature of core coolant, etc. These changes have an important impact on the results of rod ejection accident, and then affect the safety of reactor.PurposeThis study aims to verify the safety of large advanced PWR under rod ejection accident condition and the influence of key input parameters on accident analysis results.MethodsBased on the neutron dynamics software TWINKLE and fuel performance analysis program FACTRAN, the typical four types of operating conditions, including the beginning of life the hot full power and the hot zero power, the end of life the hot full power and the hot zero power, were selected to carry out the simulation calculation of the control rod ejection accident analysis for large advanced PWR, and the sensitivity analysis of key input parameters of rod ejection accident conditions was performed by using the direct numerical perturbation method.Results & ConclusionsSimulation results show that the power peak is the most sensitive to the worth of rod ejection, but less sensitive to shutdown reactivity. The consequences of the control rod ejection accident designed for CAP1400 can meet the requirements of acceptance criteria and the reactor is in the safe and controllable state.

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.

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.