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Safety is a major concern in the design, operation and development of a nuclear reactor. One aspect of nuclear reactor safety factor is thermal-hydraulics aspect. In a PWR-type nuclear power plant has been used lighter fluid coolant is water or H2O. In this research, using nanofluid Al2O3-Water with volume fraction of (1%), (2%) and also (3%), used as a cooling fluid in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement. This research was carried out modeling of fuel elements are arranged rectangular, then performed numerical simulations using Computational Fluid Dynamics (CFD) code. In order to obtain the characteristic pattern of flow velocity of each fluid, the fluid temperature distribution along the cylinder wall temperature distribution of the fuel element. Then analyzed the heat transfer in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement, including heat transfer coefficient, Nusselt number (Nu), as well as heat transfer correlations. Heat transfer correlation for nanofluid Al2O3-Water (1%), (2%) and also (3%) proved to core of PWR nuclear reactor fuel element sub channel rectangular arrangement with the Reynolds number (Re) is stretched, namely: 404 096

Abstract To protect a hermetic enclosure and the equipment and systems of a reactor installation housed in it from damage caused by the ignition (explosion) of hydrogen, most nuclear power plants with pressurized water reactors are provided with a hydrogen concentration monitoring system and an emergency hydrogen removal system. These systems prevent the formation of explosive mixtures in the accident localization zone by maintaining the concentration volume of hydrogen in the mixture below the safety limits, which ensures the preservation of the density and strength of the hermetic enclosure and the operability of other localizing safety systems. A key component of an emergency hydrogen removal system is a passive autocatalytic hydrogen recombiner that operates on the principle of the catalytic recombination of hydrogen and oxygen. There is an urgent need for a full-scale dynamic calculation of the development of emergency conditions in a nuclear power plant containment accompanied by a large release of hydrogen. In efforts to achieve this, we constructed, and justified, a simple engineering thermohydraulic model of hydrogen removal in the operation of a passive autocatalytic recombiner based on the available experimental data. This paper presents the application results of the model as a part of contour industry codes RELAP, TRACE and CORSAR, intended, among other things, for carrying out multifactor and full-scale calculations of the dynamics of emergency processes with the release of hydrogen into nuclear power plant premises. This model allows us to substantiate the dynamics of local concentrations of gas components of a mixture in a confined space; the temperature of the mixture, the catalyst and the walls of the box; and the pressure when hydrogen or steam is supplied to the box. We have analysed various rates of hydrogen supply to a closed box to numerically substantiate the time at which the concentration reaches the maximum level. Furthermore, we have calculated the performance for several entrance concentrations of hydrogen, and obtained a satisfactory agreement between the dynamics of the concentrations, the temperatures of the catalyst and gas, and the productivity of the passive autocatalytic hydrogen recombiner. These calculations are based on the results of comparisons between calculated and available experimental data. Highlights • Operation of hydrogen PAR (passive recombiners) was reviewed. • Mathematical model for PAR was developed. • Mechanism for the catalytic surface oxidation of hydrogen was reviewed. [ABSTRACT FROM AUTHOR]

All radioactive gaseous and liqjuid effluents discharged from Korean nuclear power plants are monitored by effluent monitors to prevent effluent releases to the environment under uncontrolled conditions. This paper provides the methodology and parameters used in the calculation of alarm (high) and warning (low) set-points for gaseous and liquid effluent monitors in Korean pressurized water reactors. Alarm set-points are determined to assure compliance with the Korean regulator)' limits of concentration of radioactive effluents. Even though warning set-points are not required by the regulatory body, Korean pressurized water reactors determine the warning set-points of effluent monitors not only to take an active management of effluent discharge but also to keep radiation doses to members of the public living around nuclear power plants as low as reasonably achievable. [ABSTRACT FROM AUTHOR]

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV. [ABSTRACT FROM AUTHOR]

To increase the exterior characteristics and decrease the pressure fluctuations of the coolant pump synchronously, a symmetrical-nonuniform angular repartition strategy for the vane blades was herein proposed. Firstly, for the purpose of the effective parameterization of the vane, a new parameterized method considering the blade distances and the vane install location was defined and imported. Continually, based on the parameterized method, samples in the design space were generated with the Latin Hypercube Sampling (LHS) method. Then, with the integration of the generated database, Computational Fluid Dynamic (CFD) analysis, BP_Adaboost strong classifier, an optimization strategy was finally established. Taking the scale model of CAP1400 (on a scale of 1:2.5) as the reference, the symmetrical-nonuniform angular repartition strategy was applied in the practical design process, and 3 optimal samples were gotten at last. Through comparative analysis, it could be found that the new design structures have a better energy conversion ability, and the specific descriptions are as: the optimized vanes could make the coolant pump have a better exterior performance from 0.8Q d to 1.2Q d mass flow; most importantly, the unsteady analysis quantitatively demonstrated that the optimal structures could effectively decrease the pressure fluctuations, especially at the outlet of the coolant pump. The symmetrical-nonuniform angular repartition strategy for the vane blades is expected to provide a technical reference for the third-generation nuclear power plant design. [ABSTRACT FROM AUTHOR]

This work investigates the depletion capability implemented in lattice physics code STREAM for the prediction of pressurized water reactor (PWR) uranium dioxide (UO 2 ) spent nuclear fuel (SNF) isotopic inventory. The validation of this capability is performed by comparison of STREAM calculation results to measured SNF assay data obtained from PWRs Takahama-3, Calvert Cliffs and GKN II. The depletion analysis is conducted with the ENDF/B-VII.0 library and uses a pin cell model of the fuel rods from which the fuel samples were taken. The Chebyshev Rational Approximation Method (CRAM) is used to solve the depletion equation with about 1300–1600 isotopes in the depletion chain. 16 actinides and 23 fission products are analyzed in 14 spent UO 2 fuel samples. The actinides are isotopes of uranium, neptunium, plutonium, americium and curium. The fission products nuclides include isotopes of cesium, neodymium, europium, samarium as well as 106 Ru, 144 Ce, 155 Gd, 99 Tc, 90 Sr, 109 Ag, and 103 Rh. The sensitivity of some of the nuclides to the details of the power history and the adjustment of the fuel sample burnup is discussed. The impact of using ENDF/B-VII.0 library instead of ENDF/B-VI.8 is also discussed. Most of the nuclides analyzed are well predicted within ±7% of the experiment for actinides and fission products. STREAM depletion results are also compared to the codes SWAT, HELIOS and SCALE results based on publicly available information in literature, to check the performance of STREAM relative to other codes for the prediction of SNF isotopic inventory. The comparison to other code systems shows that the implementation in STREAM is of comparable accuracy. Overall, this paper demonstrates that the depletion capability in STREAM can be reliably applied to predict the isotopic inventory of PWR UO 2 SNF for burnup ranging from 14 to 54 GWd/t and initial enrichment ranging from 3.0 to 4.1 wt% 235 U. [ABSTRACT FROM AUTHOR]

Injection of Emergency Core Cooling System (ECCS) water into the primary loops of the Pressurized Water Reactors (PWRs) leads to rapid cooling of Reactor Pressure Vessel (RPV) inside wall after Loss Of Coolant Accident (LOCA). This condition, known as Pressurized Thermal Shock (PTS) intensifies the propagation of the RPV structural defects and would be considered as an ageing mechanism. For structural and fracture analysis of RPV wall, thermal-hydraulic analysis of PTS should be accomplished to obtain the steam/water flow characteristics in the downcomer. For this purpose, simulation of steam/water stratified flow (due to density difference) after the injection point should be done by Computational Fluid Dynamics (CFD) methods. In this region, steam condensation over water layer is considered as the only heat source and controlled by turbulence eddy motion near the steam/water interface. Based on Surface Renewal Theory (SRT), Heat Transfer Coefficient (HTC) would be calculated by evaluation of turbulence length and velocity. Therefore, prediction of turbulence characteristics plays a significant role for estimation of interfacial mass transfer and temperature profile. High gradient of velocity and Turbulence Kinetic Energy (TKE), and interfacial mass and momentum transfer at the steam/water interface needs some modifications for application of traditional turbulence models. Implementation of damping function is one of the common solutions to overcome the overestimation of TKE at the steam/water interface. Although, this function improves flow characteristics of smooth stratified flow, investigation of conservation equations and experimental data implies that the other source function is needed when the flow regime changes to wavy flow. In this paper, a new source function of TKE based on variations of turbulence characteristics is proposed for steam/water interface leading to a special boundary condition of turbulence. To investigate the effects of this modification, simulation of air/water and steam/water stratified flow in three different test facilities is performed. The results show that the implementation of the source function of TKE improves the prediction of turbulence characteristics at the interface of isothermal stratified flow. Also condensation rate and temperature gradient of steam/water stratified flow have a better agreement with experimental data. [ABSTRACT FROM AUTHOR]

Abstract The paper provides a detailed methodology to perform the exergetic analysis of a Nuclear Steam Supply System (NSSS) of a Pressurized Light Water Reactor, PWR, comprised of the Nuclear Reactor and its main coolant system components (RCS). The second law analysis is carried out based on the four fundamental thermodynamic quantities: mass, Energy, Entropy and, therefore, Exergy. In particular, referring to heat exchanges within the Nuclear Reactor, the paper provides some updates of models and parameters finalized to make more and more detailed a model already developed by the authors [1]. The modeling, referring to a steady-state operational mode of the Reactor, takes into account all heat transfer phenomena between nuclear fuel UO 2 , its Zircaloy clad, cooling water, vessel material and the external environment. In the second part of the paper, the theoretical model is extended to all main RCS components (Vertical recirculating type Steam Generator, primary coolant pump and piping) to get to the Exergy Destructions and exergetic Efficiencies of all NSSS main components. A test case, applying the methodology in question, is exemplified in the Part II of the same paper. [ABSTRACT FROM AUTHOR]

Abstract The paper shows the results of the exergetic analysis of the Nuclear Steam Supply System (NSSS) of the MARS Pressurized Light Water Reactor using the theoretical methodology described in the authors' previous works [1] and [2]. The analysis firstly aims at a novel assessment of the irreversibilities occurred in the nuclear reactor vessel to compare the results, in terms of Exergy Destruction and exergetic Efficiency, with those obtained adopting one of the most employed methodology as reference. The comparison showed that a detailed exergetic analysis, mainly aimed to strictly assess the fission temperature, can lead to a higher estimate of the PWR exergetic Efficiency values. [ABSTRACT FROM AUTHOR]

The coupled environment fracture model (CEFM) has been modified and calibrated to predict crack growth rate (CGR) in Alloy 600 under typical PWR primary coolant conditions. The customized CEFM provides quantitative predictions of the effects of stress intensity factor, hydrogen concentration, yield strength, and temperature on CGR in Alloy 600 in PWR primary coolant environments, as well as explaining the dominating mechanism of stress corrosion cracking (SCC) in this alloy. The importance of the mechanical properties, such as yield strength and stress intensity factor, as identified previously from Artificial Neural Network (ANN) analysis of CGR data in the literature, has been confirmed theoretically. The success in explaining the intergranular stress corrosion cracking (IGSCC) of Alloy 600, argues that the CEFM, which was originally developed to describe IGSCC in sensitized stainless steels, is equally applicable for describing IGSCC in Alloy 600. [ABSTRACT FROM AUTHOR]

The effects of uniaxial tension properties and low cycle fatigue behavior of 16MND5 bainitic steel cylinder pre-corroded in simulated pressurized water reactor (PWR) were investigated by fatigue at room temperature in air and immersion test system, scanning electron microscopy (SEM), energy disperse spectroscopy (EDS). The experimental results indicated that the corrosion fatigue lives of 16MND5 specimen were significantly affected by the strain amplitude and simulated PWR environments. The compositions of corrosion products were complexly formed in simulated PWR environments. The porous corrosion surface of pre-corroded materials tended to generate pits as a result of promoting contact area to the fresh metal, which promoted crack initiation. For original materials, the fatigue cracks initiated at inclusions imbedded in the micro-cracks. Moreover, the simulated PWR environments degraded the mechanical properties and low cycle fatigue behavior of 16MND5 specimens remarkably. Pre-corrosion of 16MND5 specimen mainly affected the plastic term of the Coffin-Manson equation. [ABSTRACT FROM AUTHOR]

Combined cycles are used for various reasons, including to increase the efficiency of power generation system. In this study, a gas turbine cycle is combined with a pressurized water reactor (PWR) power plant to increase the total plant efficiency. In this novel cycle, saturated steam produced in the steam generators of the nuclear power plant is superheated by the hot combustion gases exiting the gas turbine. An exergoeconomic analysis is carried out and the effects of compressor pressure ratio and gas turbine inlet temperature are investigated on the net power output, the first and second law efficiencies, the total cost rate and the specific cost of the produced work. The results show that there is an optimum pressure ratio for each gas turbine inlet temperature. The combined cycle total cost rate and the specific cost of the produced work for a gas turbine inlet temperature of 1500 K and a compressor pressure ratio of 13 are determined to be 41,882 $/h and 31.63 $/MWh, respectively. [ABSTRACT FROM AUTHOR]

To investigate the exergy loss of fuel rod during converting nuclear energy into thermal energy, the partial differential equations of steady heat transfer of pressurized water reactor fuel rods and the first and second laws of thermodynamics were used. The exergy analysis method was innovatively combined with the numerical calculation of temperature field. The numerical calculation program was compiled to simulate the fuel rods and heat transfer channels and analyze the temperature distribution, the exergy loss distribution and the energy utilization efficiency during converting nuclear energy into heat energy and during coolant heat transfer. The results show that the fuel rod exergy loss is increased with latter decreasing in the axial direction and increased in the radial direction with the maximum exergy loss coefficient of 0.207 at the edge of fuel core. The exergy loss in the convective heat transfer process is mainly related to the heat transfer temperature difference and increased with latter decreasing along the thermal channel with the total exergy loss coefficient of 0.304. [ABSTRACT FROM AUTHOR]

The development of analysis models for the Swiss reactors is a key objective of the STARS project at the Paul Scherrer Institut (PSI). Within this context there is a need for the development of computational fluid dynamics (CFD) models of the Swiss reactors in support of future high fidelity investigations of steady-state and transient scenarios. This article presents initial results for the CFD analysis of a Siemens KWU PWR with a focus on the flow behaviour and heat transfer in the gap between the core shroud and core barrel. Temperatures and densities in this region of the reactor are important, for example, for accurate estimations of fast neutron fluence and activation in the steel structures of the core shroud, core barrel and reactor pressure vessel. The flow behaviour in this region may also be relevant for better understanding of ex-core detector responses. The flow conditions in the core bypass region were found to be in the transition-to-turbulence regime, with vortex shedding taking place downstream of the core formers as a result of flow instabilities. The non-stationary nature of the flow presented a challenge in terms of obtaining a solution within a reasonable time period. Two approaches were proposed to address this challenge: time-averaging of the flow-field information before solving the conjugate heat transfer problem; time-averaging of surface heat fluxes in order to derive detailed surface heat transfer coefficients. Both approaches yielded similar results with similar computational effort. Several characteristics and features of the core bypass flow are discussed. Updated Monte Carlo simulation results show that the influence of the core bypass temperatures on the neutron fluence predictions is non-negligible. This highlights the importance of including accurate bypass temperatures in future Monte Carlo simulations focused on ex-core regions. [ABSTRACT FROM AUTHOR]

Nuclear Power Plants mostly act as base-load stations mainly because of constraints on rate of reactivity addition. In order that such plants operate as commercially viable entities, it is necessary that they are capable of operating as load following stations. The inherently self-regulating nature of a Pressurized Water Reactor (PWR) makes it a natural choice for load followers. However, this entails an associated problem of periodic variation in spatial concentrations of the burnable neutron poison Xenon which alters the spatial flux profile substantially within the reactor core, and this phenomenon is known as Xenon oscillation . This paper proposes a nonlinear controller design methodology based on Nonlinear Dynamic Inversion (NDI) which is coupled with constrained optimization to develop a tracking controller that achieves load following operation of a PWR over a wide range of reactor power with no Xenon oscillations and satisfies the operational constraints of the reactor imposed by reactivity worth of the control devices and allowable fuel and coolant temperatures. [ABSTRACT FROM AUTHOR]

In order to reduce the false alarm or omission phenomenon existing in the loose parts monitoring system, reduce the possibility of that the loose parts bring great risks to the safe operation of the nuclear reactors under the impact of high speed water flow, the collision of the loose parts and the inner wall of the equipment was simulated by the collision of the six different quality steel balls and the steel plate, and then the corresponding impact signal was obtained by the acceleration sensor. The wavelet transform was used to decompose the original shock signal in the multi-layer frequency domain, and the energy ratio was used as the mass estimation feature vector. The selected eigenvector was input as a multi-layer back propagation( BP) neural network, and a suitable steel ball mass estimation model was established after training with different samples. The results showed that the estimated errors of different quality steel balls were 0.04%, 1.14%, 6.76%, 0.01%, 0.05% and 0.07%, respectively, the errors were all within 8%. By using the scientific combination of wavelet packet transform and BP neural network, the minimization of quality estimation error is realized under the control of system cost. [ABSTRACT FROM AUTHOR]

316NG austenitic stainless steel is a commonly-used material for primary coolant pipes of pressurized water reactor systems. These pipes are usually joined together by automated narrow gap welding process. In this study, welds were prepared by narrow gap welding on 316NG austenitic stainless steel pipes, and its microstructure of the welds was characterized. Then, fatigue crack growth tests were conducted at 325 °C. Precipitates enriched with Mn and Si were found in the fusion zone. The fatigue crack path was out of plane and secondary cracks initiated from the precipitate/matrix interface. A moderate acceleration of crack growth was also observed at 325°Cair and water (DO = ∼10 ppb) with f = 2 Hz. [ABSTRACT FROM AUTHOR]

The Loading Pattern Optimization (LPO) of a Nuclear Power Plant (NPP), or in-core fuel management optimization, is a real-world and prominent problem in Nuclear Engineering with the goal of finding an optimal (or near-optimal) Loading Pattern (LP), in terms of energy production, within adequate safety margins. Most of the reactor models used in the LPO problem are particular cases, such as research or power reactors with technical data that cannot be made available for several reasons, which makes the reproducibility of tests unattainable. In the present article we report the results of LPO of problems based upon reactor physics benchmarks. Since such data are well-known and widely available in the literature, it is possible to reproduce tests for comparison of techniques. We performed the LPO with the data of the benchmarks IAEA-3D and BIBLIS-2D. The Reactor Physics code RECNOD, which was used in previous works for the optimization of Angra 1 NPP in Brazil, was also used for further comparison. Four Optimization Metaheuristics (OMHs) were applied to those problems: Particle Swarm Optimization (PSO), Cross-Entropy algorithm (CE), Artificial Bee Colony (ABC) and Population-Based Incremental Learning (PBIL). For IAEA-3D, the best algorithm was the ABC. For BIBLIS-2D, PBIL was the best OMH. For Angra 1 / RECNOD optimization problem, PBIL, ABC and CE were the best OMHs. [ABSTRACT FROM AUTHOR]

• Thermophysical analysis of KLT-40S fuel assembly has been performed. • Fuel burnup values have been obtained for different dispersive fuel compositions and outer fuel rod diameters. • Kutateladze's method was applied to include effect of subcooled boiling and refine the cladding wall temperature. • Altering fuel composition to Th-U233 slightly decreases maximal temperatures of both fuel and cladding. • Increasing the outer fuel rod diameter up to 7.2 mm improves the thermophysical safety of the reactor core. This paper contains the results of KLT-40S reactor unit thermophysical studies and describes its main design features. Such quantities as fuel burnup, Reynolds and Prandtl numbers, coolant flow velocity, temperature distributions for fuel and fuel rod cladding, and departure from nucleate boiling ratio up to boiling crisis were calculated for different fuel rod diameters and fuel compositions. It is shown that changing the dispersed fuel composition from (U238 + U235)O 2 to (Th232 + U233)O 2 does not alter the maximal temperatures of fuel and fuel rod cladding significantly. An increase in the outer fuel rod diameter up to 7.2 mm leads to a decrease in the maximal temperatures of fuel and fuel rod cladding. Maximal temperatures of fuel and fuel rod cladding do not exceed melting temperatures for silumine and E635 alloy, respectively. Coolant flow velocity for all fuel rod diameters does not exceed 10 m/s and there is no boiling crisis. [ABSTRACT FROM AUTHOR]

Control design for pressurized water reactor (PWR) is difficult due to associated non-linearity, modelling uncertainties and time-varying system parameters. Extended state observer (ESO) based active disturbance rejection control (ADRC) presents a simple and robust solution which is almost model free and has few tuning parameters. However, conventional ESO suffers from noise over-amplification in the obtained estimates due to high-gain construction which in turn degrades the noise sensitivity of the closed-loop system and limits the achievable dynamic performance in practical scenarios. To overcome this problem, two recent techniques namely cascade ESO (CESO) and low-power higher-order ESO (LHESO) are implemented for control of PWR. Simulation analysis is conducted in MATLAB to illustrate the performance improvement obtained over conventional ESO based ADRC, particularly in case of time-varying disturbances. Extensive simulation analysis is also conducted to investigate robustness towards parametric uncertainties. The study also presents a comparison of conventional ESO, CESO and LHESO to highlight their advantages as well as limitations which in turn would help the users in selecting the appropriate scheme for their particular use-case. • Two noise suppressing extended state observers are designed and implemented for PWR. • Detailed stability analysis of the observers is presented for a PWR. • Detailed discussions about the noise suppression features are presented. • Extensive simulation studies are performed considering realistic scenarios. • Extensive robustness analysis is conducted considering parametric uncertainties. [ABSTRACT FROM AUTHOR]

A pressurized water reactor (PWR) is a multivariable system that consists of several subsystems such as a core, steam generator, pipings, and plenums that show highly nonlinear behavior. These components are prone to critical parameters that cause potential accidents and propagate to the entire system. Therefore, the PWR needs continuous and fast pre-accident assessment through modeling and predicting by leveraging powerful intelligence technologies. In this regard, this study demonstrates the potential of the nonlinear autoregressive with exogenous inputs (NARX) neural network modeling approach. The established NARX model is evaluated using PWR transients under perturbations in control rod velocity and core inlet coolant temperature. The trained network was then applied for state estimation, future prediction, and fault detection. The simulation results verified that the NARX model predicts the target value with good accuracy as compared with the nonlinear input-output (NIO) neural network strategy. Further, it is used to forecast the step-ahead values of the targets in the defined confidence interval and fault detection. Overall, the present study gives the benefit of the NARX approach for estimation and fault detection applications in other nuclear engineering fields. • The PWR system model was developed and simulated under perturbations. • The NARX strategy is adopted to model the PWR transients. • The produced NARX model was used for target variables prediction, future response forecast, and fault detection. • The simulation results verified that the NARX model successfully replicates the PWR transients. [ABSTRACT FROM AUTHOR]