Pressurized Water Reactor Research Articles

Prediction of Decay Heat from PWR Spent Nuclear Fuel Using Fuel Parameters

In the context of a geological repository for nuclear waste, fast and accurate predictions of decay heat are needed for different applications ranging from canister loading optimization to comparing decay heat predictions from state-of-the-art codes with experimental measurements. This work uses a large database of simulated pressurized water reactor (PWR) spent nuclear fuel (SNF) with an extensive range of fuel parameters to demonstrate that by using only the burnup, initial enrichment, and cooling time of the SNF, it is possible to predict the decay heat of a PWR SNF. A linear interpolation model has been developed using the simulated data and tested on data from decay heat measurements using a calorimeter. The model code was also made publicly available [V. Solans, “Python Script for the Prediction of Decay Heat from PWR Spent Nuclear Fuel Using Fuel Parameters,” Zenodo (2024)]. The results show that the decay heat can be well predicted, with the relative error between measurements and predictions ranging between 4% and 8%. After correcting for a systematic deviation between predictions and experimental results using the limited set of experimental measurement data available, the relative error can be further reduced to 2% to 3%.

Generating optimal reloading patterns for CNPGS Unit-3 core using multi-objective elitist teaching-learning-based optimizer

Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.

Detection of Axial Locations of Control Rods Using Movable In-core Detectors

Background Reactivity control in Pressurized Water Reactors (PWRs) relies on Rod Cluster Control Assemblies (RCCAs), or control rods (CRs), and boric acid diluted in primary circuit water. The control rods’ axial location inside the reactor core is typically provided by means of two redundant systems: the Rod Positioning Indicator (RPI) and the Rod Control System (RCS). Methods The current work focuses on the development of an additional method for the detection of control rod axial location inside the reactor core of a Pressurized Water Reactor (PWR) using the Advanced Flux Mapping System (AFMS) and its Movable In-core Detectors (MIDS). The paper presents the methodology employed and the results of the first validation campaigns. Results Two validation campaigns were carried out, targeting respectively equilibrium and non-equilibrium xenon axial distribution. The routine proved successful in identifying control rods’ axial insertion in different operating conditions including Flux Maps (FMs) acquired at 2%, 50% and 75% power. Conclusions The routine developed at Beznau Nuclear Power Plant will serve as an additional core monitoring tool to be employed in case of suspicious or faulty indication coming from the main control rod location methods. The developed method proved able to locate control rods’ axial insertion with few steps error, equivalent to maximum 1.6% of total fuel height (i.e. 5 cm as maximum error). There is room for potential improvements with regards to the uncertainty model currently implemented.

An Experimental Study of an Autonomous Heat Removal System Based on an Organic Rankine Cycle for an Advanced Nuclear Power Plant

The present study focuses on the recovery of waste heat in an autonomous safety system designed for advanced nuclear reactors. The system primarily relies on passive safety condensers, which are increasingly integrated into the design of advanced Pressurized Water Reactors (PWRs). These condensers are typically immersed in large water tanks that serve as heat sinks and are placed at sufficient heights to ensure natural circulation. Such a heat removal system can operate for an extended period, depending on the size of the tank. This research is driven by the potential to recover part of the energy stored in the boiling water volume, using it as a heat source for an Organic Rankine Cycle (ORC) system via an immersed heat exchanger. The electricity generated by the ORC engine can be used to power the system components, thereby making it self-sufficient. In particular, a pump replenishes the water tank, ensuring core cooling for a duration no longer limited by the water volume in the tank. An experimental test setup, including a boiling water pool and an ORC engine with an electrical output of approximately several hundred watts, along with an immersed evaporator, was constructed at CEA (Grenoble, France). Several test campaigns were conducted on the experimental test bench, exploring different configurations: two distinct ORC working fluids, cold source temperature variation effects, and relative positioning of the submerged evaporator and heat source within the water tank impact. These tests demonstrated the reliability of the system. The results were also used to validate both the ORC condenser and evaporator models. This article presents this innovative system, which has recently been patented. Moreover, to the best of our knowledge, the investigated configuration of an ORC that includes an immersed evaporator is original.

Unveiling the synergistic effect of thermal aging and irradiation corrosion on the passivation performance and stress corrosion cracking of 308 L stainless steel under high-temperature water environment

This study investigated coupling of thermal aging with irradiation corrosion, and carried out the microstructure and electrochemical characterization of 308 L stainless steel (SS). The passivation performance of 308 L SS was studied under the simulated pressurized water reactor (PWR) environmental condition and its stress corrosion cracking (SCC) performance was tested under thermal aging-irradiation conditions. The experimental results indicate that thermal aging results in spinodal decomposition of δ-ferrite, giving rise to G-phase precipitate and carbide precipitate formation at δ/γ-phase interface, whereas irradiation causes Cr depletion and Ni enrichment at the interface. Both thermal aging and irradiation lead to the decreased corrosion resistance of 308 L SS. Under coupling conditions of the two, severe Cr loss occurs at the phase boundary, which leads to severe corrosion on δ/γ phase boundary and also reduces SCC resistance of 308 L SS.

Design a robust intelligent power controller for pressurized water reactor using particle swarm optimization algorithm

Design a robust intelligent power controller for pressurized water reactor using particle swarm optimization algorithm

Indirect neutron radiography experiment on dummy nuclear fuel rods for pressurized water reactors at CMRR

Indirect neutron radiography experiment on dummy nuclear fuel rods for pressurized water reactors at CMRR

Estimation of Turbulent Mixing Factor and Study of Turbulent Flow Structures in Pressurized Water Reactor Sub Channel by Direct Numerical Simulation

Abstract Subchannel analysis codes are presently a requirement for design and safety analysis of nuclear reactors. Among the crucial inputs for these codes, the turbulent mixing factor holds particular significance. However, acquiring this factor through experimental means proves to be a challenging endeavor, primarily due to the necessity for precise pressure equilibrium between subchannels. Consequently, this requirement leads to the undertaking of expensive and intricate experiments for each new reactor or in cases where there are modifications in fuel bundle design. The need for direct numerical simulation (DNS) stems from the challenges and costs involved in experimental techniques, and the uncertainties due to empiricism in computational fluid dynamics (CFD) models. In this study, DNS has been conducted across six Reynolds numbers, ranging from 17,640 to 1.176 × 105, in the geometry of a pressurized water reactor (PWR) subchannel. The resulting turbulent flow structures have been computed and their dynamics are examined. Furthermore, this paper presents a methodology for directly calculating the turbulent mixing factor from the fluctuating velocity field obtained from DNS data. The turbulent mixing process has been scrutinized in-depth, and a correlation for the turbulent mixing factor is developed. It is noted that most of the mixing occurs in the near-wall region. The study suggests different mixing factors for mass and momentum mixing. This paper aims to provide a comprehensive insight into the turbulent mixing phenomenon.

TSUNAMI Analyses of the Similarity and Applicability of Zero Energy Deuterium 2 Critical Experiments for Reactor Physics Code Benchmarking for a Pressurized Water Reactor Small Modular Reactor Design Concept

Abstract This paper presents the findings of studies conducted at Canadian Nuclear Laboratories (CNL) to support the development of small modular reactor (SMR) designs. The primary focus of this research was to evaluate the suitability of the zero energy deuterium 2 (ZED-2) critical facility in replicating the reactor physics environment for a pressurized water reactor small modular reactor (PWR-SMR) design concept through similarity and nuclear data sensitivity studies, using the TSUNAMI code suite. It was found that previous ZED-2 experiments would be quite promising for application to a PWR-SMR design. Further similarity and sensitivity studies of hypothetical mixed-lattice substitution experiments, where PWR-SMR fuel assemblies were placed into a substitution region of the ZED-2 critical facility demonstrated improved similarity. Subsequent analyses focused on the impacts of dissolved Gadolinium (Gd) and boron (B) neutron absorbers, suggesting the feasibility of using future ZED-2 experiments to more closely replicate PWR-SMR reactor physics behavior. Building on these initial findings, the design for PWR-SMR fuel assembly substitution experiments in the ZED-2 facility were explored further. These hypothetical experiments feature water-cooled PWR-type fuel assemblies inside a shroud, surrounded by heavy-water-moderated CANdu FLEXible, Low Enriched Uranium, Recovered Uranium (CANFLEX-LEU/CANFLEX-RU) fuel channels. Similarity and sensitivity studies indicate a very high level of similarity of these experiments for PWR-SMR design applications.

A subchannel analysis code for advanced liquid metal fast reactor cores and study on heat transfer characteristics of core geometry parameters

The liquid metal fast reactor represents an important reactor type for the future development of nuclear energy. Accurately modeling and predicting the subchannel flow and heat transfer phenomenon in the rod bundles plays a key role in ensuring core safety. Consequently, a subchannel code suitable for the liquid metal fast reactor is analyzed based on the subchannel analysis code of the Pressurized Water Reactor (PWR). The modified code incorporates various types of liquid metals, such as lead, lead-bismuth eutectic (LBE), and sodium, along with their constitutive models, enhancing the applicability of the liquid metal reactor systems. This code has been verified and validated at several levels, including comparing the code simulation results with other similar subchannel codes results, high-dimensional Computational Fluid Dynamics (CFD) simulations, and experiment data. The sensitivity analysis of core geometric parameters and turbulent mixing coefficients is performed based on the modified version code. The influence of the core geometry, including the rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) and the wire wrap pitch (H) on the sensitivity of outlet temperature has been investigated. The results show that the new subchannel analysis code suitable for liquid metal cooled reactor shows reasonable agreement with the verified and validated results. The geometric parameters, such as P/D, RTW and H have noticeable effects on the outlet temperature difference of the subchannels. Additionally, the outlet temperature distribution in the subchannel is significantly affected by different turbulent mixing coefficients and the distribution of the turbulent mixing coefficient also influenced by the reactor core sizes. Overall, this study could support the rod bundles design and safety analysis in the liquid metal reactors.

Performance of a hydraulic buffer for PWR fuel assemblies: Mathematical modeling, numerical solutions, and experimental comparison

In a Pressurized Water Reactor (PWR), the hydraulic buffer serves as an essential component, significantly mitigating the impact force between the control rod drive mechanism and the fuel assembly in scenarios of emergency shutdown. This paper provides a complete analysis of the dynamical performance of a new type of hydraulic buffer, including its mathematical modeling, numerical solution scheme, and experimental comparison. During the fluid modeling, five typical cases are first classified in terms of both flow directions and coefficients based on the relative positions of the sleeve and the piston. A new flow iterative calculation format encompassing flow directions that can efficiently solve the flow coefficients is proposed during the fluid modeling. A one-way coupling scheme is used for the fluid–structure dynamical solutions. An experimental comparative study is conducted using the standard and reference (SR) case, and the present method shows good agreement with the experiment. In the present analysis: (1) The dynamic characteristics of the buffer are fully demonstrated using time history curves and phase diagrams, the physical parameters of fluid and structural motions, mainly including fluid pressure inside the chamber, dynamic relative displacement and velocity of the piston and sleeve, the rebound disparagement, and system kinetic energy; (2) Typical features of double peaks of the impact force, related to the collision between the piston, the impacted object, and the sleeve, have been captured and well simulated; (3) The variation of the two peaks of the impact forces with parameters and the transformation laws of the maximum impact force between these two peaks have been fully revealed.

Study on flow regime characteristics in 3×3 rod bundle channels based on wire mesh sensor

Abstract The characteristics and evolution of two-phase flow are of critical importance to the safety of pressurized water reactors. Based on the double-layer wire mesh sensors in this paper, the air-water two-phase flow experiment of the 3×3 rod bundle channels is carried out at room temperature and pressure. The results show that the critical bubble diameter range for the reversal of lateral lift direction is 4 to 5.8 mm. In addition, the time-averaged void fraction for bubbly flow reveals a wall peak distribution at lower superficial gas velocities and shifts to a core peak as these velocities increase. For cap flow, the cross-distribution of cap bubbles within adjacent subchannels triggers large-scale mixing of the liquid phase between adjacent subchannels. For slug flow, large-sized bubbles develop along the axis and cross subchannel gaps to aggregate into slug-shaped bubbles, with a more pronounced distribution of the central peak of void fraction. The drift-flux models are evaluated against the experimental data, and the Ozaki model demonstrates higher precision in estimating void fraction, exhibiting a deviation of 9.8% on average.

Evolution of fretting wear behavior of zirconium alloy cladding tube under gross slip regime in simulated primary water of pressurized water reactor

The evolution of fretting wear behavior of zirconium alloy cladding tubes mated with dimples under the gross slip regime (GSR) was investigated. The findings revealed that the primary wear mechanisms under GSR were delamination, surface fatigue wear and abrasive wear, and the fretting damage rate mainly depends on delamination. The cross-sectional microstructure of the worn area could be divided into the third-body layer, tribologically transformed structure layer, and general deformation layer, with their formation mechanisms analyzed. Furthermore, the mechanism of wear-induced grain refinement was identified as dynamic recrystallization (DRX), including both continuous DRX and discontinuous DRX. Additionally, the processes of fretting wear and DRX were discussed.

Risk-informed design: An unavailability allocation approach

Due to the technological complexity of nuclear power plants and the large number of components involved in their design, the objective space for balancing nuclear safety, availability performance, and cost in an optimization model becomes extremely vast. Consequently, this poses challenges in achieving operational optimization at the plant level, complicating the assurance of compliance with regulatory requirements. One solution is to partition the problem and analyze it at a system level to reduce the objective space. In this work, different options are analyzed to optimize cost and unavailability for systems that provide the means for meeting previously proposed unavailability targets. This ensures that licensing basis events (LBE) comply with risk targets and regulatory criteria. Additionally, the role of system unavailability due to maintenance activities and reliability issues related to unplanned component failures (e.g., random failures, common cause failures) in the optimization problem is analyzed. The various proposed optimization methods are implemented in the design of a pressurized water reactor (PWR) system. Evolutionary algorithms are used as optimization methods, with Genetic Algorithms for single-objective problems and Non-dominated Sorting Genetic Algorithm III (NSGA-III) for multi-objective problems. The main finding suggests that traditional models, which aim to minimize costs and unavailability, face challenges in adapting to the systems unavailability target. Therefore, optimization occurs by shifting the focus towards minimizing costs and distance to the unavailability target. This approach ensures that the results closely align with the unavailability target, thereby creating more efficiency in operating and maintenance costs while also ensuring acceptable design and operational regulatory compliance.

Hoop tensile properties and crack propagation investigation of 2D braided SiCf/SiC composite tubes: Experiments and simulations

SiCf/SiC composites are promising candidates for advanced pressurized water reactors (PWRs) fuel cladding materials due to their enhanced accident tolerance. Their mechanical properties are strongly influenced by the braided structure of continuous SiC fibers. This study fabricated SiCf/SiC composite tubes with braid angles ranging from 30° to 50°, evaluating their hoop tensile properties through expansion-due-to-compression (EDC) experiments and analyzing the damage process using finite element method simulation. Results indicate that variations in braid angles significantly affect structural density, thereby impacting mechanical strength under hoop tensile stress. Increased braid angles result in smaller pore units and higher pore density, leading to local stress concentrations and varied deflections at overlapping regions. The dynamic propagation behavior of cracks was investigated through acoustic and structural nondestructive testing methods. Finite element analysis of different braid configurations highlights the pivotal role of pore units in the initiation and propagation of hoop tensile cracks. This study enhances the understanding of toughening structures in 2D braided composites and provides a theoretical basis for future accident-tolerant fuel cladding design.

Application of the PSI cycle check-up methodology for full-core Monte Carlo simulations verified using pin power estimates and validated for hot zero power conditions

Monte Carlo (MC) based neutron transport codes provide high-fidelity numerical solutions to problems beyond the modelling capabilities of conventional core simulation methods. However, performing core-follow calculations with MC codes remains challenging due to the necessity of incorporating thermal-hydraulic feedback and generating evolved fuel composition for the cycle of interest by taking into account the entire irradiation history of loaded fuel assemblies over previous cycles. Nevertheless, at PSI, the neutronic version of a cycle check-up methodology (CHUP) is under development to address this challenge. This methodology involves extracting operating conditions and nuclide compositions from validated reference deterministic core-follow models (CASMO5/SIMULATE5/SNF), subsequently generating MC neutron transport models for selected operating points. This article presents the verification and validation performed for a hot zero power operating condition of a Swiss pressurised water reactor using the updated CHUP methodology, which automates information extraction from reference core models to minimise human intervention. The MC models were verified by assessing their deviation from criticality, consistently found to be supercritical within the [10,60] pcm range. Additionally, radial relative power distribution verification yield deviations in the [−5,4] % range compared to validated nodal results. Finally, additional validation was performed using start-up measurements from two successive cycles. Eight MC models were found to be subcritical, on average by 51 pcm, with a standard deviation of 25 pcm.

Machine Learning-Assisted Prediction of Stress Corrosion Crack Growth Rate in Stainless Steel

Stainless-steel is extensively utilized in the key structural components of the main equipment in the nuclear island of pressurized water reactor nuclear power plants. The operational experience of nuclear power plants demonstrates that stress corrosion is one of the significant factors influencing the long-term safe operation of stainless steel in the high-temperature water of pressurized water reactor nuclear power plants. This study is based on the stress corrosion crack growth rate data of 316SS and 304SS stainless steel in the simulated primary water environment of pressurized water reactor nuclear power plants. Data mining and modeling were conducted using multiple machine learning algorithms, including Random Forest (RF), eXtreme Gradient Boosting (XGBoost), Support Vector Regression (SVR), and Gaussian Process Regression (GPR), and the Sharpley Additive explanation (SHAP) method was employed to analyze the interpretability of the model. The results indicate that the stress corrosion crack growth rate prediction model based on XGBoost outperforms other models in all assessment indicators. Compared with empirical equations, XGBoost exhibits high flexibility and excellent data-driven learning capabilities. In the test set, 90% of the prediction errors are within the range of experimental values, with the maximum error multiple being 2.5, which significantly improves the prediction accuracy. Moreover, the distribution of SHAP values is consistent with the theoretical study of the stress corrosion behavior of stainless-steel, effectively reflecting the impact of cold working, temperature, and stress intensity factor on the stress corrosion crack growth rate, thereby proving the reliability of the model’s prediction results. The achievements of this study hold significant reference value and application prospects for the prediction of the stress corrosion behavior of stainless-steel in a high-temperature and high-pressure water environment of pressurized water reactor nuclear power plants.

Depressurization strategy of China's advanced pressurized water reactor under typical accidents

Advanced pressurized water reactor was simulated using system analysis program. The depressurization strategy for the automatic depressurization system (ADS) under typical ADS-triggering accidents (2-inch/10-inch cold leg break, inadvertent opening of the ADS, DVI pipeline break) was investigated. The results show that after these accidents, ADS can ensure the operation of the three types of safe injection tanks, with high redundancy and inherent safety. The multi-stage design of ADS aids the sparger in maintaining a stable critical jet state. In 10-inch cold leg break accident, the sparger cannot reached the critical jet state under all of pipeline fault conditions. From a system point of view, adjusting the pipeline design of ADS is necessary to avoid long-term condensation oscillation of the sparger in some accidents. From a design perspective of the local equipment, the nozzle structure of the sparger should attenuate the dynamic load caused by the condensation oscillation and back-attack phenomenon.

On the Feasibility of a RGA-Based Decoupling Control in the Reactor-Follow-Turbine Operation of a PWR

Closed-loop control of the reactor-follow-turbine operation is attempted in this work. The reactor power is intended to follow the turbine load demand, whereas the coolant average temperature is to be maintained constant for reasons of design safety. To this effect, a standardized pressurized water reactor model is adopted wherein the reactor external reactivity and the feedwater mass flux are deemed the manipulated control signals. Offline system identification is carried out on the model input/output (I/O) terminals to extract first order plus time delay models thereof. A further relative gain array analysis is conducted to determine the most tightly coupled I/O channels. A proportional integral controller is thereafter designed for each channel, making use of an analytical gain/phase margin tuning mechanism. The results for different output tracking scenarios confirm a feasible and quite satisfactory closed-loop control practice.

Flexible operation of nuclear hybrid energy systems for load following and water desalination

Nuclear hybrid energy systems (NHES) have the potential to provide dependable and emission-free electricity to the grid while also increasing the flexibility and reliability of the electrical grid. Molten salt reactor (MSR) technology can provide consistent, carbon-free electricity while also increasing efficiency, security, and sustainability and reducing nuclear waste. This study investigates the integration of Molten Salt Reactors (MSR) and conventional Pressurized Water Reactors (PWR) with desalination technologies: Direct Contact Membrane Distillation (DCMD), Multi-Stage Flash Distillation (MSFD), and Reverse Osmosis (RO). Dynamic first-principles models were developed and tested using real grid data from the New York Independent System Operator. The results demonstrate that nuclear power is capable of flexibly responding to changing grid demand while simultaneously producing clean water, particularly during periods of low electricity demand. The MSR-RO system was found to be the most efficient in electricity generation and water production, and all hybrid systems reduced CO2 emissions by 356,000 to 682,000 tons annually. Economic analysis reveals that nuclear desalination technologies are cost-competitive with conventional systems, especially when paired with RO. These findings confirm the technical feasibility and environmental benefits of nuclear hybrid systems for sustainable electricity and water production.