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Nuclear power plants in the United States are increasingly challenged to compete in wholesale electricity markets due to the low electricity costs and increasingly dynamic grid conditions from competing generation sources. An alternative market for nuclear power is industrial facilities that can use the thermal and/or electrical power generated by a nuclear power plant to offset the economic losses incurred by electricity market challenges. A generic pressurized water reactor (PWR) simulator was used to show the results of a basic design for a generic thermal power extraction system and tests were run using a set of procedures to show what happens when a nuclear power plant transitions from full electrical power dispatch to 15% and 50% thermal power dispatch. This type of operation leads to losses in turbine performance efficiency due to the deviation from the design operating point, but because the thermal power is also used by the industry load without conversion losses, the combined thermal efficiency of the PWR increases. For the 15% case, the thermal efficiency increased from 32% to 41.9%, while for the 50% case, the efficiency increased up to 60.1%. In addition, for 50% thermal power dispatch, the power dissipated by the condenser decreased from approximately 2000 to approximately 1300 MW (thermal), indicating a substantially diminished impact on the environment in terms of releasing heat into the cooling water reservoir.

All radioactive gaseous and liquid 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 regulatory 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.

Featured Application: This work shows a global review with three analyses of the PWR nuclear design: (i) technical evolution; (ii) level of implementation in the world, and (iii) life extension scenario. Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%. [ABSTRACT FROM AUTHOR]

Nuclear steam generators (NSGs) are key components of nuclear power plants (NPPs) since their reliability affects the overall plant performance. Corrosion control of steam generators under operation and during plant outages is attained by proper design, adequate material selection and control of water chemistry in the entire secondary cooling circuit. This article reviews the types of steam generators, the materials used for their construction (with a special focus on tubing alloys), the corrosion processes that may occur under operation and the strategies to preserve steam generators and other components of the secondary cooling system from corrosion during plant outages. [ABSTRACT FROM AUTHOR]

To study different functionalities of the PWR‐900 MW(e) reactor core, an electronic model of Simulation Program with Integrated Circuit Enfesied (PSPICE) for electronic circuit design was adopted to model the mathematical models. A classical breadboard electronically simulates the physical terminal characteristics of each network. The electronic simulator, parity simulator, for reactor core can simulate six‐delayed neutron groups, single prompt neutron, neutron source, fuel channel, the coolant, reactor poisons, temperature feedback, control rods, safety rods, and boron injection. Positive/negative step or sin reactivity functions are two case studies were analyzed. The obtained results were found to be in a good agreement with the available data of El‐Dabaa PWR 900 MW(e) basic simulator. [ABSTRACT FROM AUTHOR]

The corrosion products in the primary loop of the nuclear reactor can be deposited on the surface of the fuel rod, forming a deposition layer called CRUD layer, introducing the heat transfer resistance. The CRUD layer will result in an increase in the temperature of the cladding and accelerating the corrosion of the cladding. In this work, the corrosion product deposition and thermal resistance models of the deposition layer were established. The deposition behavior of corrosion products on the surface of a novel Helical-Cruciform Fuel (HCF) assembly and its influence on fluid-solid conjugate heat transfer were studied by coupling the deposition models with the flow and heat transfer models for fuel rods and coolant. The results show that the deposition rate of particles is affected by the wall shear force; the deposition rate of impurity ions is affected by the wall shear force and the temperature of the coolant in the viscous layer. The erosion rate of the deposited layer is related to the wall shear force and the sum of the particle and ion deposition rates. The thickness of CRUD layer on the surface of HCF assembly increases with the decrease of coolant flow rate and the increase of fuel power. Furthermore, the CRUD layer thickness varies in a period of 90° along the circumference of the fuel and reaches a peak at the root of the fuel ribs. [ABSTRACT FROM AUTHOR]

In pressurized water reactor (PWR) nuclear plants, reactor pressure vessel (RPV) nozzles are critical due to geometric stress concentrations that worsen with large temperature gradients. In the context of Atucha-I Long Term Operation (LTO) assessment, this work presents a high-fidelity thermo-mechanical simulation approach to verify the stress levels on the main nozzles of the vessel under a pressurized thermal shock event. Nucleoeléctrica Argentina S.A. (NA-SA) has provided the boundary conditions, including the thermal plume that results from computational fluid dynamics (CFD) simulations of the transient event. Then, using a one-way and non-conformal projection procedure, CFD data is coupled to a highly-detailed 3D finite element model (FEM), including the RPV, closure head, nuts and bolts, and moderator tank, all interconnected through non-linear contact condition that accounts for friction. Finally, a 3D FEM submodel assuming an elastoplastic material behavior with non-linear isotropic hardening is used to verify the structural integrity of the nozzle. Results indicate that the plastic strain and stresses on the nozzles are low enough to ensure the correct function of the reactor during the LTO period. • Thermo-mechanical verification of the main nozzles of the Atucha I pressure vessel. • The highly-detailed 3D FEM model of RPV and internals uses contact with friction. • Temperatures, pressures, and a thermal plume computed with CFD were considered. • Both linear elastic and elasto-plastic material models have been compared. • Plastic strain and stresses on the main nozzle ensure safe function of the reactor. [ABSTRACT FROM AUTHOR]

Nuclear power has garnered significant attention as a clean energy resource in expediting carbon peaking and achieving carbon neutrality. It contributes to sustainable economic growth while addressing national energy security concerns. As one of the most crucial components, pressurized water reactors require improvements in several key aspects, including strengthening the natural circulation capacity, increasing the reactor power, and reducing the reactor volume. However, the volume of pressurized water reactors is usually positively correlated with their power. Increasing the reactor power results in an expansion of its volume, which poses challenges for the reactor layout and also requires additional investment. Hence, in order to reduce the reactor volume while meeting its performance requirements, this study introduces a novel design optimization approach utilizing an improved quantum fish swarm algorithm and intuitionistic linguistic decision-making. Firstly, a nonlinear mathematical model for volume optimization of the steam generator (SG) and pressurizer (PR) is constructed. Then, the improved quantum fish swarm algorithm is proposed to obtain volume optimization alternatives for the SG and PR. Quantum computation strategies are introduced to enhance the global optimization ability and population diversity. Additionally, the vision range and step size of the quantum fish swarm are adjusted adaptively to improve the optimization precision. To further accelerate the algorithm convergence, the prospect-elimination strategy is built. Last but not least, the intuitionistic linguistic decision-making is presented to gain the optimal design alternative by taking four characteristics into account, namely, the fuzziness and hesitancy of evaluation information, the consensus reaching process, the bounded rationality of decision-making behaviors, and the complementarity, independence and redundancy among nuclear evaluation criteria. A case study on the SG and PR design optimization is implemented to demonstrate the reliability and reasonability of the proposed approach. [ABSTRACT FROM AUTHOR]

This work outlines an approach for localizing anomalies in nuclear reactor cores during their steady state operation, employing deep, one-dimensional, convolutional neural networks. Anomalies are characterized by the application of perturbation diagnostic techniques, based on the analysis of the so-called “neutron-noise” signals: that is, fluctuations of the neutron flux around the mean value observed in a steady-state power level. The proposed methodology is comprised of three steps: initially, certain reactor core perturbations scenarios are simulated in software, creating the respective perturbation datasets, which are specific to a given reactor geometry; then, the said datasets are used to train deep learning models that learn to identify and locate the given perturbations within the nuclear reactor core; lastly, the models are tested on actual plant measurements. The overall methodology is validated on hexagonal, pre-Konvoi, pressurized water, and VVER-1000 type nuclear reactors. The simulated data are generated by the FEMFFUSION code, which is extended in order to deal with the hexagonal geometry in the time and frequency domains. The examined perturbations are absorbers of variable strength, and the trained models are tested on actual plant data acquired by the in-core detectors of the Temelín VVER-1000 Power Plant in the Czech Republic. The whole approach is realized in the framework of Euratom’s CORTEX project.

Cooling system in nuclear reactor is much influenced by counter-current flow. When a leakage is occured, coolant in primary circuit of hotleg and hot steam from reactor will flows in counter-current condition. An experimental investigation has been performed to examine the effect of flooding phenomenon on the pressure drop. It was observed using acrylic complex pipe as hotleg simulator consisting of three parts: horizontal, bend and inclined pipe. Geometrical sizes of pipes was in scale of 1/30 of the actual hotleg geometrical size in the PWR with inner diameter=25.4 mm, the horizontal pipe length L=635 mm, and inclined pipe length was 20 mm with an angle of 50o. Pressure drop was meassured by using differential pressure sensor installed in the lower-tank (RPV simulator) dan upper-tank (SG simulator). Pressure drop was observed before, during and after flooding. Based on the observation, it was found that the change of flow pattern is followed by the change of pressure drop. The results of this study were also compared with some correlations using the dimensionless superficial velocity parameter by Wallis correlation. Wallis correlation can be written as JG*1/2 + 0,3974 JL*1/2 = 0,4832 .

An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Nuclear Power Plants (NPPs) based on Pressurized Water Reactors (PWRs) technology are considered an alternative to fossil fuels plants due to their reliability with low operational cost and low CO2 emissions. An example of PWR plant is Angra 2 built in Brazil. This NPP has a nominal electric power output of 1300 MW and made it possible for the country save its water resources during electricity generation from hydraulic plants, and improved Brazilian knowledge and technology in nuclear research area. Despite all these benefits, PWR plants generally have a relatively low thermal efficiency combined with a large amount of irreversibility generation or exergy destruction in their components, reducing their capacity to produce work. Because of that, it is important to assess such systems to understand how each component impacts on system efficiency. Based on that, the aim of this work is to evaluate Angra 2 by performing energy and exergy analyses to quantify the thermodynamic performance of this PWR plant and its components. The methodology consists in the development of a mathematical model in EES (Engineering Equation Solver) software based on thermodynamic states in addition to energy and exergy balance equations. According to the results, Angra 2 has energy efficiency of 36.18% and exergy efficiency of 49.24%. Reactor core is the most inefficient device in the NPP; it has exergy efficiency of 67.16% and is responsible for 63.88% of all exergy destroyed in Angra 2.

In the two-step method for nuclear reactor simulation, lattice physics calculations are performed to compute homogenized cross-sections for a variety of burnups and lattice configurations. A nodal code is then used to perform full-core analysis using the pre-calculated homogenized cross-sections. One source of uncertainty introduced in this method is that the lattice configuration or depletion conditions typically do not match a pre-calculated one from the lattice physics simulations. Therefore, some interpolation model must be used to estimate the homogenized cross-sections in the nodal code. This current study provides a methodology for sensitivity analysis to quantify the impact of state variables on the homogenized cross-sections. This methodology also allows for analyses of the historical effect that the state variables have on homogenized cross-sections. An application of this methodology on a lattice for the Westinghouse AP1000® reactor is presented where coolant density, fuel temperature, soluble boron concentration, and control rod insertion are the state variables of interest. The effects of considering the instantaneous values of the state variables, historical values of the state variables, and burnup-averaged values of the state variables are analyzed. Using these methods, it was found that a linear model that only considers the instantaneous and burnup-averaged values of state variables can fail to capture some variations in the homogenized cross-sections.

Highlights • CRSE operations under MSHIM control are simulated for the whole cycle life of the AP1000 reactor. • Dynamic characteristics of the reactor during CRSE transients are analyzed with feasibility of the CRSE logic demonstrated. • Trigger rod position in the CRSE logic is optimized with the performance and effectiveness of the optimization assessed. Abstract The purpose of this paper is to study dynamic characteristics of the AP1000 reactor during control rod sequence exchange (CRSE) operations under the Mechanical Shim (MSHIM) control strategy so as to assess the feasibility of and further optimize the CRSE logic. For this purpose, a dynamic simulation platform was developed in MATLAB/Simulink environment first, with implementation of a one-dimensional nodal core model and the reactor power control system integrating the basic MSHIM control logic and the CRSE logic. Based on the platform, simulation studies of typical CRSE transients of the AP1000 reactor at the beginning of cycle life (BOC), middle of cycle life (MOC) and end of cycle life (EOL) have been performed to determine the performance of the current CRSE design. From the simulation results, impacts of the CRSE operations on the reactor power and axial power difference have been analyzed. It shows that the AP1000 reactor can accomplish the CRSE operations smoothly and safely implementing the CRSE logic. To further improve the control performance of the reactor during CRSE operations, the trigger rod position, a key factor in the CRSE design, has been optimized with the optimization results assessed through dynamic simulations against the above CRSE transients. The simulation results have demonstrated the effectiveness of the optimization. [ABSTRACT FROM AUTHOR]

All licensable critical heat flux (CHF) correlations/regressions models must determine and demonstrate a "design limit" that bounds the CHF correlation predicted/measured residuals via a 95/95 tolerance limit. This is a quick and straightforward calculation when the residuals are well behaved, exhibiting no trends and no heteroscedasticity. However, as models become increasingly complex and as required parameter ranges become more extended, the likelihood of nonconservative subregions increases. A suggested solution from the open literature is the overly conservative approach of basing the design limit on the subregion with the largest variance. This approach unavoidably overly constrains the overall regression model and often is too conservative for subregions due to a loss in degrees of freedom. Quantile regressions alleviate these issues by smoothly varying the design limit based on covariates and adapting to each subregion. Thus, a quantile regression achieves the objective of appropriately bounding all subregions without overly biasing the overall regression model. [ABSTRACT FROM AUTHOR]

Thermodynamic evaluation of a nuclear desalination plant by coupling a pressurized water reactor and a multi-effect distillation system was performed by using a new version of the Desalination Thermodynamic Optimization Program. For energy management of a nuclear desalination plant, all possible coupling configurations between the nuclear power plant and the desalination process were investigated. The software was upgraded to study exergy analysis by adding related exergy equations. Several important output parameters, such as power loss ratio, net efficiency, thermal utilization, and total power requirement, were compared; these parameters varied from 8.7% to 28.5%, 32.6% to 33.3%, 36.1% to 36.9% and 14.7 to 35.3 MWe, respectively. The minimum power loss ratio and the maximum net efficiency and thermal utilization were obtained from multi-extraction points, which included the heat rejected by the condenser. Exergy analysis showed that the maximum exergy destructions (1,292.9 MWth) were related to the reactor core, which is considered the main irreversible component. The results demonstrated good agreement with previous theoretical and experimental studies. [ABSTRACT FROM AUTHOR]

The constraint annealing method is presented and demonstrated for the solution of large-scale, multiconstrained problems in light water reactor fuel cycle optimization. Constraint annealing is a penalty-free method that eliminates the need for traditional constraint weighting factors by treating each objective function and constraint as separate and concurrently solved minimization problems within a global optimization search framework. The current application seeks to demonstrate the effectiveness of constraint annealing for a complex core loading pattern design problem containing multiple objective functions and constraints without the need for additional ad hoc control parameters. Two problems of varying degrees of complexity are analyzed. The first problem is defined by a single objective function based on maximizing cycle energy with two constraints based on power peaking and peak rod exposure. The second problem expands upon the first by adding an additional objective function for vessel fluence and four additional constraints based on controlled power peaking, steaming rate, moderator temperature coefficient, and alternate source term. Results demonstrate that constraint annealing inherently addresses issues of scaling associated with different objective function and constraint formulations as well as the impact on cycle energy. [ABSTRACT FROM AUTHOR]

The surfaces of 316L stainless steel with mechanical polishing (MP) or electropolishing (EP) in the presence of a magnetic field (MEP) before and after exposure to simulated pressurized water reactor (PWR) primary water were characterized by various techniques such as SEM, Raman spectroscopy, XPS and TEM measurements. The surface microhardness was lower and the surface roughness was higher for the as-treated MEP specimen than for the as-treated MP specimen. The oxide films that formed on the as-treated MP and MEP surfaces were nanometer-thick and rich in chromium. The Cr/(Fe + Ni) ratio of the as-treated MEP surface film was higher than that of the as-treated MP surface film. After exposure to simulated PWR primary water at 310 °C for 1050 h, a dark double oxide layer that was approximately 662 nm thick formed on the MP surface, while the MEP surface retained its original brightness as before exposure with a nanometer-thick oxide layer with a high Cr content. The MP and MEP specimens showed different oxidation kinetics when exposed to high temperature water. The high Cr/(Fe + Ni) ratio of the surface film on the MEP 316L SS contributed to its excellent oxidation resistance in simulated PWR primary water. • 316L SS treated by magneto-electropolishing (MEP) under a magnetic field. • About 2 nm thick oxide film on as-treated mechanically polished (MP) and MEP surfaces. • A double layer sub-micrometer thick oxide on exposed MP surface in PWR water. • A nanometer-thick oxide of high Cr content on exposed MEP surface in PWR water. • High Cr/(Fe + Ni) ratio of MEP surface contributed to its excellent oxidation resistance. [ABSTRACT FROM AUTHOR]

Abstract High-temperature water corrosion tests were performed to investigate the corrosion behaviors of Alloy 690 TT in simulated PWR secondary water with gaseous hydrogen permeation condition. Double-layer oxide films formed on hydrogen-free and hydrogen-permeating specimens. The thickness of a Cr-rich inner oxide film on the hydrogen-free specimens has no obvious difference from the hydrogen-permeating ones. For hydrogen-free specimens, the outer oxide films were Fe and Ni enriched spinel-type oxide particles. For hydrogen-permeating specimens, outer oxide films were sheet-like or needle-like Cr-rich chromia-type oxides. In summary, the solid diffusing hydrogen affected the morphology and composition of outer oxides. Highlights • High temperature water corrosion test was performed in hydrogen permeation condition. • The thickness of Cr-rich inner films has no obvious change by hydrogen permeation. • Sheet-like outer oxides on hydrogen-permeating samples unlike hydrogen-free ones. • Fe and Ni-rich spinel-type outer oxide films formed on hydrogen-free specimens. • Cr-rich chromia-type outer oxide films formed on hydrogen-permeating specimens. [ABSTRACT FROM AUTHOR]

Highlights • New, improved matched index of refraction material combination for 5 × 5 rod bundle. • Full field PIV measurements throughout entire test section. • Full field uncertainty quantification of ensemble-averaged and fluctuating velocity. • Novel method of vector elimination in lieu of explicit masking. • Data at and below spacer grid mixing vane tips. Abstract A new combination of materials with matched indices of refraction was used in average and statistical PIV analysis of a 5 × 5 rod bundle with prototypical pressurized water reactor (PWR) mixing vane spacer grids. The new combination of materials shows improved matching of refractive index which enabled detailed PIV measurement of subchannel flow, even when viewed through a relatively large number of material interfaces. Convergence of average full field PIV measurements at ∼ 27 , 000 Reynolds number was demonstrated as independent of measurement plane depth relative to camera view window. A method for average vector post-processing elimination is proposed using validation statistics from PIV. The proposed method shows potential for eliminating vectors in areas of poor-convergence for planes with little interfering geometry. A full uncertainty quantification was performed on the full field of each PIV measurement planes. Similar to convergence rates, uncertainty demonstrated independence from measurement plane depth. Additional material interface between camera and measurement plane does not appear to have affected instantaneous velocity measurements' contribution to average vector fields. Symmetric measurement planes were compared between 2 rod pitch intervals of the spacer grid. Velocity and statistically derived quantities show good coherence between symmetric planes. Measurements near and even below the tips of the mixing vanes were collected and analyzed. [ABSTRACT FROM AUTHOR]

Highlights • A time-dependent simulation mode has been implemented in the Serpent 2 code. • The reactor physics modeling of the SPERT transients have been performed using Serpent 2. • The results of the reactor physics simulations without fuel temperature feedback are in very good agreement with the experimental data. Abstract Monte Carlo multi-physics modeling of reactor transients is being made feasible by the advancement of computer capacities. Developments have recently been made in the area of dynamic Monte Carlo methods, and corresponding new algorithms have been implemented in a number of transport codes. In particular, a time-dependent simulation mode has been implemented in the Serpent 2 code. This paper is focused on time-dependent simulations of the SPERT experiments. The results of reactor-physics simulations are presented. The modeling has been performed using Serpent 2. The results of the reactor physics simulations without fuel temperature feedback are in very good agreement with the experimental data. [ABSTRACT FROM AUTHOR]

Highlights • Verification of STREAM capabilities to predict spent fuel radiation source terms. • Validation of decay heat calculations using 91 decay heat measurements for 52 PWR fuel assemblies. • Results show good agreement between calculated and measured decay heat. Abstract This paper presents the verification and validation of the radiation source term calculation capability implemented in the reactor analysis code STREAM for pressurized water reactor (PWR) spent nuclear fuel (SNF) analysis. Activity, decay heat, neutron and gamma source spectra of irradiated PWR fuel assemblies are calculated with STREAM and ORIGEN and the results obtained are compared. The verification is performed with Westinghouse 15 × 15, 17 × 17 and Combustion Engineering 16 × 16 designs with enrichment of 3–5 wt.% 235U, burnup of 30–60 GWd/t and cooling times of 0.3–1000 years. The radiation source term comparison between STREAM and ORIGEN indicates that the implementation of the source term equations is correct. The validation is done by comparing STREAM calculation results against 91 decay heat calorimetric measurements for 52 PWR fuel assemblies performed at the Swedish central storage facility for spent nuclear fuel, CLAB, and United States Hanford Engineering Development Laboratory (HEDL) and General Electric Morris Operations. The measured fuel assemblies have enrichment of 2–4 wt.% 235U, burnup of 28–51 GWd/t and cooling times of 2–27 years. The comparison between calculated and measured decay heat shows good agreement. For most of the measurements analyzed, the calculated decay heat is within the range of the measurement uncertainty. The average, overall C/E over 71 CLAB decay heat measurements is 1.000 ± 0.017, and 0.997 ± 0.013 for 20 US measurements. Overall, this work demonstrates that STREAM can be reliably used to predict PWR SNF radiation source terms. [ABSTRACT FROM AUTHOR]

Countercurrent flow limitation (CCFL) in a pressurized water reactor hot-leg pipe geometry with a 190-mm pipe diameter was investigated experimentally and numerically at the COLLIDER test facility of the Technical University Munich in the past 3 years. This paper summarizes the most important CCFL findings learned from the COLLIDER test facility and tries to explain the reasons for obtaining different descriptions, results, and conclusions at different CCFL experimental investigations. The factors that can affect CCFL experimental results are explained in detail including some scale effects. The necessary preconditions to compare two sets of data from different CCFL experimental investigations are discussed in detail. The difference among CCFL-related limits/curves is clarified taking data at the COLLIDER as an example. The limits included the limit of the transition from a supercritical into a subcritical flow (SSTL); the onset of CCFL limit (iCCFL) inside the hot-leg pipe; the onset of CCFL limit (eCCFL) at the entrance of the steam generator; the deflooding limit (CCFLd); the CCFL characteristics curve (CCFLch), which predicts the water delivery rate after the onset of iCCFL; and the onset of hysteresis limit. It will be shown that among these limits only SSTL, CCFLch, and eCCFL are original limits while the rest are derivatives of them. In particular, it will be shown that the iCCFL limit is a combination of the SSTL and CCFLch limits. The effect of scale upon the eCCFL's mechanism (whether a water accumulation or droplet entrainment at the entrance to the steam generator) is clarified via a comparison to a 50-mm CCFL facility at Kobe University. This paper tests the scalability of interface distribution at quasi-stationary conditions (i.e., points along the CCFLch curve) via a comparison of time-averaged interface distributions obtained at similar inlet conditions ( at the COLLIDER 190-mm and Kobe 50-mm channels. The comparison will show that interface distributions (which are directly linked to the pressure drop and interfacial momentum transfer) cannot be scaled at the bend/riser/entrance region because of the influence of the channel diameter upon occurring CCFL mechanism. Meanwhile, the water level gradient can be similar at the horizontal part, but not the relative water depth. [ABSTRACT FROM AUTHOR]

Fuel-coolant interaction (FCI) is an important issue for the assessment of severe accident safety for both sodium-cooled fast reactors (SFRs) and pressurized water reactors (PWRs). For the ASTRID SFR demonstrator, FCI is a key phenomenon affecting the relocation of molten fuel in engineered discharge tubes between the core region and the core catcher plenum. FCI controls jet fragmentation and debris bed formation and raises the issue of potentially energetic vapor explosions in the ASTRID lower head. In this frame, experimental data will be necessary to validate SCONE, the fuel-sodium interaction code under development at CEA. For PWRs, one of the configurations of interest lies within the residual case where in-vessel retention would fail. In this case, it is expected that a light metallic layer would be the first to interact with water, before a heavier oxide melt discharge. Here, steam explosion and debris bed formation are the two major points of interest. Based on the experimental expertise gained from the KROTOS facility and its X-ray radioscopic imaging system, new test facilities have been designed to carry out prototypic (depleted uranium-containing) corium interactions with either sodium or water in PLINIUS2, the CEA future large-mass experimental platform dealing with masses above 100 kg. Some test sections have been specially designed to ensure proper visualization of the fuel, liquid coolant, and vapor phases by an improved X-Ray imaging system. This paper presents the future PLINIUS 2 platform as well as the experimental programs foreseen to study both water-corium and sodium-corium interactions. [ABSTRACT FROM AUTHOR]

Highlights • A nonequilibrium nonlinear three-region pressurizer model was proposed and linearized. • Transfer function models of pressurizer during insurge and outsurge transients were developed for controller design. • Pressure and water level controllers of a small PWR pressurizer were designed based on the transfer function models. • Simulation studies show effectiveness of the pressurizer models and robustness of the designed controllers. Abstract A pressurizer is a key equipment to ensure the safe operation of a pressurized water reactor by maintaining the reactor coolant system pressure within allowed tolerances. In this paper, a nonequilibrium three-region pressurizer model for the pressurized water reactor was developed first base on the basic conservation laws of mass and energy for the steam and water in the pressurizer. Then the nonlinear model was linearized to introduce the transfer function models of the pressurizer during insurge and outsurge transients for the controller design of a small pressurized water reactor pressurizer. Based on the developed transfer function models, the closed-loop pressurizer pressure and water level control systems including conventional proportional-integral-derivative controllers were designed employing the control strategy that uses a spray valve and two electric heaters for pressure control and regulates the charging flow rate with the letdown flow rate keeping constant for water level control. Finally, a simulation platform for the small pressurized water reactor pressurizer was developed in MATLAB/Simulink with implementation of the proposed nonlinear and nonequilibrium three-region pressurizer model, a widely-used nonlinear and nonequilibrium three-region pressurizer model and the designed pressure and water level controllers. Two typical load change transients, including the 100% to 90% of full power step load decrease and load rejection from 100% to 25% of full power, were simulated based on the platform. Dynamic responses of the pressure and water level obtained using the two nonlinear pressurizer models were analyzed, and the controller performances were assessed. The analysis and assessment results have shown satisfactory control performance and good robustness of the designed controllers regardless of the pressurizer simulation model adopted, demonstrating the feasibility, effectiveness and accuracy of the developed nonlinear and transfer function models of the pressurizer for dynamic simulation and controller design. [ABSTRACT FROM AUTHOR]

The Generic Pressurized Water Reactor (GPWR) simulator has been used in the Nuclear I&C Laboratory at Khalifa University (KU) since 2013 to improve student performance in nuclear engineering that is a multidisciplinary field involving nuclear reactor physics, thermodynamics, fluid mechanics, thermal hydraulics, radiation, etc. The simulator, developed by Western Service Corporation, has been integrated as a teaching and educational tool in different Engineering Programs at KU (Mechanical and Nuclear engineering). This lab is used in an undergraduate course where students apply the knowledge taught from different courses such as nuclear systems, fuel cycle, thermal hydraulics, safety principle, and control functions through a virtual operating NPP simulator. This real-time, full scope and high fidelity simulator allows to perform different operating conditions such as plant startups, shutdowns, and load maneuvers; as well as normal and abnormal plant transients, and critical scenarios and accidents. Since its installation in the Nuclear I&C Laboratory at KU in 2013, thirty students have benefited from this learning simulator. The main skills and learning outcomes expected to be achieved by students through the using of this tool are (i) ability to describe different NPP components and understand different process occurring in different subsystems, (ii) explain and apply safety principles and protective protocols, and (iii) analyze and interpret the plant behavior during transient operations and when severe accidents happen.

Employing a linear reactivity burnup-dependent model, we investigate how the new axial flattened power densities (Pontes and Driscoll models) affect the steady-state axial temperature distributions of the sub-channel in pressurized water reactors (PWR). Two approaches were used: analytical, using mean parameters, and numerical, using appropriate empirical correlations. The Pontes distribution showed advantages in all evaluated regions except in the centerline of the fuel rod, outperformed by the Driscoll distribution. The results indicate that the linear reactivity model (LRM) is extremely advantageous in the evaluated reactor, as it reduces axial temperatures on hot spots and has the potential to enhance the reactor's safety, efficiency, and longevity. • The LRM profiles significantly reduce maximum temperatures on the channel. • A constant power function has limitations in areas such as cladding and coolant. • Driscoll' Model reduced the maximum temperatures on the cladding and pellet. • Pontes' Model reduced the maximum temperatures at all analyzed regions. • The new models enhanced the efficiency and safety of PWR reactors. [ABSTRACT FROM AUTHOR]

• Use of uranium mononitride fuel in PWR. • Neutronic and Thermal Hydraulic analysis using SCALE 6.0 and STHIRP codes. • Simulation of a typical PWR fuel assembly to compare the physical parameters of UO 2 and UN fuels. • The largest thermal conductivity of uranium mononitride contributes to lowest fuel temperature. • The UN fuel presents highest production of transuranic elements and considerable 14C production during the fuel burnup. The nitride fuel is currently considered as a material to replacement of uranium dioxide in Light Water Reactors (LWR), mainly due to its higher thermal conductivity when compared with the oxide fuels and similar melting point. This behavior may contribute to larger heat conduction in the fuel assemblies and contributes to increase the safety margins to fuel melting during operation. In this way, the present work aims to contribute to studies on the use of uranium mononitride (UN) in LWRs. Neutronic and thermal hydraulic analysis were performed for a typical assembly of Pressurized Water Reactor (PWR) using uranium dioxide (UO 2) and uranium mononitride (UN) fuels. The SCALE 6.0 (Standardized Computer Analyses for Licensing Evaluation – version 6.0) and STHIRP (Thermal Hydraulic Simulation of Research Reactors) codes were used to calculate the neutronic and thermal hydraulic parameters, respectively. The goal is to verify the possible use of UN in a PWR. Comparing the evaluated fuels, the largest thermal conductivity of UN contributes to largest heat conduction in the pellet zone and to lowest fuel temperature. However, the UN presents highest production of transuranic elements and considerable 14C production into UN during the burnup. Results of neutronic and thermal hydraulic parameters behavior for both fuels are presented in details in this work as well as carefully analyzed. [ABSTRACT FROM AUTHOR]

The feasibility and the limitations of using nuclear energy as a dispatchable power source for covering the daily fluctuations of the solar electricity are examined. In particular, the perspective of electricity production in Slovenia until 2050 is investigated, focussing on the projected rapid increase in the solar electricity production and the resulting large-scale annual and daily fluctuations. However, the study is relevant for any electrical grid where photovoltaics is likely to become one of the main sources of electricity. A simulation study based on a nonlinear pressurized water reactor (PWR) model with 2-point neutron kinetics controlled by two groups of control rods using a new simplified control approach is presented. The relevant nonlinearities and the controllability issue affecting this control configuration are discussed. It is shown that a nuclear power plant can be used to compensate for some of the expected power fluctuations using historical data of the solar power production in Slovenia. • Nuclear power is considered for covering the fluctuations of photovoltaic electricity production. • Electricity production in 2020 and its projection until 2050 in Slovenia are explored. • A pressurized water reactor is simulated with a 2-point nonlinear model. • A new simplified control approach with two sets of control rods is presented. • The nonlinearities and the controllability issue affecting control are explained. [ABSTRACT FROM AUTHOR]

The use of Machine Learning (ML) has gained significant attention in recent years for use in the development of surrogate models capable of predicting various neutronics parameters in Pressurized Water Reactors (PWRs). These approaches aim to provide fast and accurate approximations of conventional model outputs which are typically both computationally expensive and time-consuming to run. A novel supervised learning framework LatticeNet , employing Artificial Neural Networks (ANNs) and computer vision techniques, has been shown to be particularly useful in the prediction of assembly and pin-level parameters such as multiplication factor or pin powers. This framework was validated using deterministically-generated training data for a standard fuel assembly, representing best-case input conditions. This paper investigates the performance of LatticeNet as well as the tools developed alongside when given Monte Carlo-generated inputs and assemblies of varying pin dimensions, assessing the network's tolerance to training data uncertainty and atypical configurations. [ABSTRACT FROM AUTHOR]

• Robust and accurate prediction of PWR safety parameters is provided by TCNN. • Sparse connections greatly improve the global awareness of the mode. • Convolutional kernels effectively enhance local feature extraction ability. • Basic blocks consist of residual connections prevent the gradient loss. Accurate forecasts for pressurized water reactor safety parameters are essential to ensure the safe operation of nuclear reactors. Potential of artificial neural networks on this task is limited owing to the lack of extracting the core location information. Sparse connections have unique advantages in discovering correlation between neighboring components and convolution kernels are designed to deal with two-dimensional information. In this paper, topological information embedded convolutional neural network (TCNN) was firstly established and utilized. This model enhanced the ability of fusing location features and component attributes through sparse connections and convolution layers. Datasets of China's Qinshan Nuclear Power Plant Phase II PWR was used to evaluate the performance of TCNN. Comparative and ablation experiments demonstrated that TCNN has superiority in working as efficient predictor for pressurized water reactor safety parameters, indicating that the proposed model promoted the digitalization of nuclear power plants. [ABSTRACT FROM AUTHOR]

• A Monte Carlo simulation has been conducted for various dispersed fuel compositions of reactor KLT-40S. • Use of (Th232 + U233)O 2 fuel composition increases fuel lifetime by 450 EFPD and fuel burnup by 45 MW∙d/kg HM. • (Th232 + U233)O 2 fuel composition at 7.2 mm diameter results in a 45% fuel lifetime increase and a 5% fuel burnup decrease. • (Th232 + U233)O 2 composition requires less volume of moderator at fuel rod diameters of 7.2 mm and 8.2 mm. • All fuel compositions, except for (Th232 + U233)O 2 , show optimal fuel cycle parameter values at project fuel rod diameter. This paper demonstrates the results of the Monte Carlo simulation conducted with MCU-PTR software for a small modular pressurized water reactor KLT-40S. Four dispersed fuel compositions, (U238 + U235)O 2 , (U238 + Pu239)O 2 , (Th232 + U235)O 2 , and (Th232 + U233)O 2 , were examined for fuel lifetime and fuel burnup calculations. The transition to the (Th232 + U233)O 2 fuel composition extends fuel lifetime by 50% and fuel burnup by 44% at design value of the fuel rod diameter. If the fuel rod diameter for (Th232 + U233)O 2 fuel composition is increased to 7.2 mm, fuel lifetime is extended by 105%, and fuel burnup is extended by 38% compared to the design values. [ABSTRACT FROM AUTHOR]

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV). For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%.