Pressurized Water Reactor Research Articles

Large eddy simulation of a turbulent flow in a 5 x 5 rod bundle with a mixing vane spacer grid

In fuel assemblies of pressurized water reactors (PWRs), mixing-vane spacer grids are extensively employed to array fuel rods and enhance inter- and intra-subchannel mixing. The turbulent flow downstream of spacer grids is essential knowledge to support the design of spacer grids and mixing vanes. As a high-fidelity approach for turbulence simulation, large eddy simulation (LES) results are assessed with the high-resolution flow field measured with particle image velocimetry (PIV). The LES results reveal a mixing vane effect on vortex structures and Reynolds stresses. The mechanism of vortex generation at the mixing vanes and the development downstream of the spacer grid are visualized by the Q-criterion. The distributions of the root mean square fluctuating velocity and Reynolds stresses decay rapidly downstream of the spacer grid. The secondary flow intensity downstream of the spacer grid predicted by LES is evaluated and follows an exponentially decaying law with distance from the spacer grid.

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

Actinide minor is a reactor waste with high toxicity and a long half-life. Minor actinides can be reduced by reusing them as fuel mixtures in reactors. This research uses PWR reactors with the primary fuel UN-PuN or Uranium Plutonium Nitride with a burning time of 5 years. The fuel consists of enriched Uranium, reactor-grade Plutonium from LWR waste, and minor actinides including Neptunium-237, Americium-241, and Curium-244. The purpose of this study was to find a design that is effective in reducing minor actinide waste. There are six designs or cases used in the addition of minor actinides. Each case has six minor actinide pins in each assembly. The addition of minor actinides is arranged in heterogeneous cores. The analysis was carried out by observing the values of k-eff, excess reactivity, and mass of minor actinides obtained from simulations using OpenMC code 0.13.2 and the ENDF/B-VIII library. The homogeneous core obtained an excess reactivity of 9.7 % with a percentage of plutonium of 8 %. The results of the homogeneous core are used as a reference for preparing a heterogeneous core. The heterogeneous core obtained an excess reactivity of 9.9 % with a percentage of plutonium F1: 5.5 %, F2: 8 %, and F3: 10.5 %. Np-237 can be reduced by 53 kg, and Am-241 can be reduced by 61 kg with minor actinide pins in case 1. Cm-244 can be reduced by 363 kilograms with minor actinide pins in case 6. Excess reactivity in the addition of Np-237 and Am-241 decreased to 5.3 %, while the accumulation of Cm-244 increased to 12.1 %.

An XPS and TEM study of the composition and structure of native oxides on the inner surface of as-received Ni base alloy steam generator tubes

Ni based alloys steam generator (SG) tubes in pressurized water reactors may suffer generalized corrosion and Ni release in primary water. The processes of corrosion and oxidation are strongly dependent on the surface properties. In particular, the chemical composition and structure of the native oxide films present at the surface of the SG tubes are crucial and improving the passive properties of the surface could contribute to reduce the corrosion damage and Ni release. The native oxides on the inner surface of two as-received Ni base alloy SG tubes were investigated using two main techniques, namely X-ray photoemission spectroscopy and transmission electron microscopy. The chemical elements present in the native film for the two tubes were identified and the structure and composition of the native film at the nanoscale was determined. A duplex oxide structure was identified for both samples, but with differences in the chemical composition of the inner and outer oxide layer. Chemical heterogeneities such as alumina, Ti oxides and Cr oxides were also detected; they are considered to play a role in the evolution of the surface exposed to primary water. The two techniques used allowed a multi-scale description of the surface and led to complementary results.

Development and Testing of a System Thermal-Hydraulics Model for a 50-MWel-Class Pressurized Water Reactor–Small Modular Reactor

Abstract This paper describes the development, analysis, testing of a RELAP5-3D system thermal-hydraulics model for a 50-MWel-class pressurized water reactor–small modular reactor (PWR–SMR), similar to that by NuScale Power. This study focuses on a series of sensitivity tests to investigate the impacts of model changes. Parameters considered in the sensitivity study included the surge line junction resistance (SLJR), steam generator (SG) heat transfer area (SGHTA), SG primary flow area (SGPFA), SG secondary pressure (SGSP), and SG secondary flowrate (SGSF). Results for the reference and sensitivity simulations are compared with available design data. The flow in the primary circuit of the PWR–SMR is driven by natural circulation and can be sensitive to changes in hydraulic resistance and pressure drop in system components. Initial analysis results demonstrated significant flow oscillations. As a result of sensitivity studies, it was found that the surge line junction resistance needed to be increased by increasing the form loss coefficient from 4.9 to 30.0, to reduce mass flow oscillation amplitude to less than ±2%. Modifications to the steam generator heat transfer area, primary flow area, or secondary pressure have very little impact in reducing flow oscillations. However, it was found that the steam generator secondary flowrate will affect primary circuit flow oscillations, and when the SGSF was artificially increased from 68 kg/s (design data) to 91 kg/s (a 36% increase), the oscillations were eliminated, along with better matching with design data for core flowrate and inlet/outlet temperatures. Further improvements to the model for the steam generator will be required.

Effect of heavy ion irradiation on the electrochemical behavior of 321 stainless steel

Aiming to the environment for 321 stainless steel (SS) in primary circuit of pressurized water reactor (PWR), the effect of heavy ion irradiation to simulate neutron irradiation on the electrochemical behavior of alloy is unclear. The purpose of this work is to evaluate the corrosion resistance of 321 SS at different doses and the contribution of heavy ion irradiation in it. This work reveals that the corrosion resistance decreases in the order of 2 dpa > 0 dpa > 10 dpa > 60 dpa at room temperature, consistent with the sequence of δ-phase content near the surface. The effect of irradiation dose on the corrosion resistance of 321 SS is inversely proportional at higher dose and optimal performance of the SS irradiated to 2 dpa is attributed to the highest content of Cr-rich δ-phase. The pits occur in C enrichment zone, accompanied by increase of O content, while no pit occurs in Ti enrichment area. The results provide a reference for the irradiated SS with heavy ion to acquire optimal pitting resistence.

Burnup-Dependent Neutron Spectrum Behaviour of a Pressurised Water Reactor Fuel Assembly

Understanding the behaviour of a neutron spectrum with burnup is important for describing various phenomena associated with reactor operation. The quest to understand the neutron spectrum comes with a lot of questions. One question that is usually asked by students is: Does the neutron spectrum harden or soften with burnup? Most textbooks used by students do not provide a definite answer to this question. This paper seeks to answer this question using a 3D model of a standard 17 × 17 pressurised water reactor fuel assembly. Two cases were studied using the Serpent Monte Carlo code: the first considered the fuel assembly with constant boron concentration (traditionally found in many published papers), and the second considered boron iteration (where the boron concentration was reduced with burnup). Neutron spectra for the two cases at beginning of life and end of life were compared for spectral shifts. In addition, thermal spectral indices were used to assess spectrum hardening or softening with burnup. Spectral shifts to lower energies were observed in the thermal region of the neutron spectrum, whereas the fast region experienced no spectral shift. There was an increase in thermal spectral indices indicating that the spectrum became soft with burnup.

Optimal controller design for reactor core power stabilization in a pressurized water reactor: Applications of gold rush algorithm.

Nuclear energy (NE) is seen as a reliable choice for ensuring the security of the world's energy supply, and it has only lately begun to be advocated as a strategy for reducing climate change in order to meet low-carbon energy transition goals. To achieve flexible operation across a wide operating range when it participates in peak regulation in the power systems, the pressurised water reactor (PWR) NE systems must overcome the nonlinearity problem induced by the substantial variation. In light of this viewpoint, the objective of this work is to evaluate the reactor core (main component) of the NE system via different recent optimization techniques. The PWR, which is the most common form, is the reactor under investigation. For controlling the movement of control rods that correspond with reactivity for power regulation the PWR, PID controller is employed. This study presents a dynamic model of the PWR, which includes the reactor core, the upper and lower plenums, and the piping that connects the reactor core to the steam alternator is analyzed and investigated. The PWR dynamic model is controlled by a PID controller optimized by the gold rush optimizer (GRO) built on the integration of the time-weighted square error performance indicator. Additionally, to exhibit the efficacy of the presented GRO, the dragonfly approach, Arithmetic algorithm, and planet optimization algorithm are used to adjust the PID controller parameters. Furthermore, a comparison among the optimized PID gains with the applied algorithms shows great accuracy, efficacy, and effectiveness of the proposed GRO. MATLAB\ Simulink program is used to model and simulate the system components and the applied algorithms. The simulation findings demonstrate that the suggested optimized PID control strategy has superior efficiency and resilience in terms of less overshoot and settling time.

Modeling particle deposition in the primary circuit of pressurized water reactors for the OSCAR code

Particulate corrosion products are generated in the primary system of pressurized water reactors (PWRs) by volume precipitation and by erosion of oxides formed on metal surfaces through their uniform corrosion. The activation of corrosion products, mainly 58Co and 60Co (respectively coming from the activation of 58Ni and 59Co) leads to radiation field growth around the primary system, directly impacting system integrity and the radioprotection of nuclear workers. In order to understand and mitigate contamination by activated corrosion products, contamination predictions can be performed using the OSCAR code, which relies on the development of models to describe the numerous and complex interactions at stake. Particulate corrosion products account for a significant portion of corrosion products, as such the deposition/erosion mechanisms have their importance for the overall computation of surface or volume contamination. The aim of this article is to present an updated and inclusive deposition model for particulate corrosion products by taking into account surface interactions. The impact of the new deposition model on contamination predictions is then evaluated and has enabled to reproduce, for the first time using the OSCAR code, the preferential contamination in 58Co in the cold side of the circuit, measured by gamma spectrometry with the EMECC device on commercially PWRs.

Understanding Flow Fouling Deposition and Solute Hideout-Return Behavior at the Phase Change Interface.

Nuclear energy is a competitive green energy, yet corrosion deposition and boron hideout on pressurized water reactor fuel cladding surfaces could cause localized corrosion and power shift, resulting in huge safety and economic risks. Alleviation of these problems requires the understanding of the corrosion deposition mechanism and related boron behavior. In this study, we explore corrosion product deposition in typical fuel assembly channels under subcooled boiling conditions and propose a boron hideout and return mechanism to explain the reason for the failure of the power reduction inhibiting a power shift. Porous corrosion depositions with the same morphology and thickness as the real depositions in a fuel cycle are obtained in a week via the accelerated deposition method simulating a real subcooled boiling and water chemical environment. Stronger subcooled boiling generates more bubbles, resulting in higher supersaturation of corrosion products at the gas-liquid interface. The corresponding precipitated stable crystals are smaller, and the formed deposition layer is looser and thicker with smaller particles. On the basis of the above characterizations, the effect of subcooled boiling, solute concentration, and water chemistry on the corrosion deposition mechanism is revealed. High-resolution characterization methods indicate that boron hides within the depositions mainly in the form of H3BO3 and Li2B4O7. The boron coolant concentration increases by 307.9 ppm after power reduction, confirming the return behavior of porous hidden boron. Hidden boron return behavior brings potential challenges for estimating critical conditions and plant response operations. The results of this study provide a precise method for understanding the corrosion product deposition and boron hideout-return behavior to further develop mitigation strategies for power shift and localized corrosion security issues.

Stress Corrosion Cracking of High-Strength Ni-Base Alloys in Pressurized Water Reactor Primary Water Containing KOH vs. LiOH

The stress corrosion cracking (SCC) behavior of high-strength Ni-base Alloys X-750 and 718 has been investigated to assess the effect of replacing LiOH with KOH as the pH moderator in pressurized water reactor (PWR) primary circuit coolant. The SCC initiation behavior of X-750 was evaluated by multitensile specimen constant load test in either LiOH- or KOH-containing PWR primary water. The SCC growth behavior of X-750 and Alloy 718 was evaluated on compact tension specimens with on-the-fly changes between LiOH- and KOH-containing water in three different PWR primary water chemistries. Direct current potential drop technique was used for in situ monitoring of crack initiation time and crack extension in SCC initiation and propagation tests, respectively. The results suggest no significant effect of KOH vs. LiOH on the SCC initiation or propagation of the tested materials.

Development of a heat transfer model for the film boiling regime in a fouling layer

Fouling accumulation on fuel claddings and steam generators of pressurized water reactors can significantly affect boiling heat transfer. The present study proposes a heat transfer model for the film boiling regime in a fouling layer. The proposed model assumes that heat transfer in the fouling layer occurs in three modes: boiling, liquid convective heat transfer within the fouling and at the coolant-foulng interface. The thickness of the steam film within the fouling is determined by the pressure balance method. The prediction error of the present model for wall superheat is about 20%. The sensitivity analysis results show that the pore size significantly influences heat transfer, and the wall superheat can vary by up to 200K within the maximum pore diameter range of the fouling layer. However, chimney tilt angle has a negligible effect on heat transfer since the pressure drop in the steam chimney accounts for only a small proportion.

Highly detailed structural integrity assessment of the reactor pressure vessel nozzle of Atucha-I during a pressurized thermal shock event

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.

Practical considerations for the adoption of Anderson acceleration in nonlinear diffusion accelerated transport

The use of the Anderson Acceleration (AA) method within the context of the coarse mesh finite difference (CMFD)-accelerated neutron transport k-eigenvalue calculation is explored for practical problems. Current application of CMFD only has a linear convergence rate. The AA method is similar to a quasi-Newton method and can improve the convergence rate of fixed-point iterations. We therefore implement AA for the 2D calculation in MPACT and study its performance in pressurized water reactor (PWR) problems. The initial results show that the traditional AA method can improve the convergence rate in a trivial problem and improve the stability of the CMFD method for some problems. However, the success of AA in the practical problems, where simulations with symmetry or domain decomposition, relies on one essential factor–the treatment of iterative angular flux boundary conditions in the AA. Additionally, the efficiency of AA is also affected by the computational cost for solving the optimization problem, and the computational cost of the transport solver and the CMFD solver. To address these issues, we propose a new implementation of AA that considers only the mutligroup scalar flux in the optimization problem, but the angular flux and the scalar flux in the mixing. The new approach achieves better stability and efficiency in practical simulations, and can be explained by the theory of Randomized Numerical Linear Algebra (RandNLA).

NuScale-like SMR Model Development and Applied Safety Analyses with the Code Chain Serpent-DYN3D-ATHLET

NuScale is an integral pressurized water reactor (iPWR) operated with light water driven by natural circulation through two helical coil steam generators (HCSGs). This work reports the safety analyses of a boron dilution and steam line break accidental sequences in a developed plant computational model based on the specifications from the Final Safety Analysis Report (FSAR). Multi-physics calculations are performed with the code chain Serpent-DYN3D-ATHLET. A state-of-the-art multi-dimensional vessel topology is developed with ATHLET for the accurate representation of flow and temperature fields, as well as spatial core power distribution. The static calculation results show agreement with the reference values from the FSAR. The boron dilution sequence shows a homogeneous core power excursion by the boron feedback, and the reactor is tripped upon “high pressurizer pressure” signal. During the steam line break sequence, the affected HCSG depressurizes rapidly and the reactor is tripped upon “low main steam pressure” signal. None of the transients violate safety margins. The adopted 3-D vessel modelling approach and applied multi-physics calculations are able to capture both transient physics within the reactor domain, and conclude that symmetric arrangement of the HCSG tubes enhance coolant mixing and prevent a heterogeneous core power excursion.

Fuel depletion study of the molten salt demonstration reactor

In this study, a molten salt reactor model and a pressurized water reactor model were run in SCALE for comparison. The parameters of interest in the analysis were the concentrations of different nuclides in a molten salt reactor, specifically compared with concentrations of these same nuclides in a pressurized water reactor. This paper discusses this analysis, as well as discussing and comparing the burnup in both models. The results suggest that using a mechanistic source term for radionuclides instead of a bounding, conservative source term enables more realistic reactor modeling, making it a great option for licensing both advanced reactors and light-water reactors.

Short communication: “Effect of dissolved hydrogen on the early corrosion behavior of 316 stainless steel in simulated PWR environments”

In simulated pressurized water reactor environments, the corrosion test on 316 stainless steel was conducted for 40 to 4000 s, with 25 and 50 cc/kg-H2O dissolved hydrogen concentrations (NH and HH). It was found that the external Fe-rich oxide particles and the inner Cr-rich layer were simultaneously formed at the beginning of corrosion. However, the first formed Cr-rich layer was mainly Cr(OH)3 in the NH condition, while Cr2O3 in the HH condition. As exposure time increased, fraction of Cr2O3 increased due to transformation from Cr(OH)3 in both conditions. In addition, for HH condition, partial transformation to FeCr2O4 was observed.

Effectiveness re-evaluation on the intentional primary system depressurization during Zion-like Westinghouse PWR station blackout considering pressure dependence of radionuclides release

Previous studies have successfully resolved the issue of high-pressure melt ejection (HPME) followed by direct containment heating (DCH) during a total station blackout of Zion-like Westinghouse pressurized water reactor (PWR). This resolution is crucial, as the earliest occurrence of hot-leg creep failure can cause a decrease in the pressure difference between the reactor pressure vessel (RPV) and the primary containment vessel (PCV), falling below the cut-off pressure for HPME/DCH. As a recommended accident management strategy, intentional depressurization of the reactor coolant system (RCS) has been proposed. Depressurization leads to a delayed accident progression due to accumulator injection and maintains a pressure difference at the time of RPV failure below the cut-off pressure. In the present analyses using MAAP5, it was observed that depressurization before core heat-up, achieved by opening 2 power-operated relief valves (PORVs), resulted in the most delayed RPV failure, consistent with previous studies. However, present sensitivity analysis considering the pressure-dependent release of radionuclides from the fuel revealed significant changes in both the accident progression and the point of creep failure. In addition, the creep-failure point shifted from the RPV bottom to the RPV sidewall, and the pressure difference at that location exceeded the cut-off pressure. To prevent sidewall failure of the RPV, it may be advisable to employ a depressurization rate higher than that achieved by using 2 PORVs, even if it slightly accelerates the accident progression.

Scaling analysis for the condensation induced water hammer effect in horizontal pipe

In nuclear power engineering, researchers pay significant attention to phenomena related to the condensation process involving direct contact between saturated steam and subcooled water, mainly the phenomenon known as Condensation Induced Water Hammer (CIWH). Within the steam pipes of the pressurized water reactor, when saturated steam comes into direct contact with subcooled water, it may form an isolated steam pocket surrounded by water. The condensation of steam within this pocket results in its subsequent collapse, accelerating the surrounding water column to collide with the pipe wall and causing a substantial pressure peak. In this paper, the scaling analysis was conducted on the CIWH phenomenon in the steam line of a small PWR. The research is rooted in a two-fluid, six-equation framework and employs equation-based analysis to investigate the non-dimensional number of CIWH occurring within horizontal pipelines. By analyzing the conservation equations, suitable non-dimensional criteria for thermohydraulic phenomena were determined. These criteria can be employed to ascertain experimental facilities' geometric dimensions and operating conditions, which can study the risks associated with CIWH phenomenon.

An Innovational Jacobian-Split Newton-Krylov Method Combining the Advantages of the Jacobian-Free Newton-Krylov Method and the Finite Difference Jacobian-Based Newton-Krylov Method

The Jacobian-free Newton-Krylov (JFNK) method is a widely used and flexible numerical method for solving the neutronic/thermal-hydraulic coupling system. The main property of JFNK is that the Jacobian-vector product is evaluated approximately by finite difference, avoiding the forming and storage of Jacobian explicitly. However, the lack of an efficient preconditioner is a major bottleneck for the JFNK method, leading to poor convergence. The finite difference Jacobian-based Newton-Krylov (DJNK) method is another advanced numerical method, in which the Jacobian matrix is formed and stored explicitly. The DJNK method can provide a better preconditioner for Krylov iteration than JFNK. However, how to compute the Jacobian matrix efficiently and automatically is a key issue for the DJNK method. By fully utilizing the sparsity of the Jacobian matrix and graph coloring algorithm, the Jacobian can be computed efficiently. Unfortunately, when there are dense rows/blocks, a huge computational burden will emerge due to the lack of sparsity, resulting in the extremely poor efficiency of Jacobian computation. In this work, a Jacobian-split Newton-Krylov (JSNK) method is proposed to resolve the dense row/block problem by combining the advantages of JFNK and DJNK. The main feature of the JSNK method is to split the Jacobian matrix into sparse and dense parts. The sparse part of the Jacobian matrix is explicitly constructed using the graph coloring algorithm while for the dense part, the Jacobian-vector product is approximated by finite difference. The computational complexity of the JSNK method is analyzed and compared to the JFNK method and the DJNK method from theoretical and experimental aspects and under different meshes. A simplified two-dimensional (2-D) high-temperature gas-cooled reactor (HTR) model and a simplified 2-D pressurized water reactor model are utilized to demonstrate the superiority of the JSNK method. The numerical results show that the JSNK method successfully resolved the dense rows/blocks. More importantly, its efficiency significantly outperforms the JFNK method and the DJNK method.

Heat transfer optimization of uo2-mo fuel using genetic algorithms

Two genetic algorithm (GA) methods were applied to thermal finite element models to optimize the heat transfer efficacy of a UO2-Mo composite fuel pellet with typical pressurized water reactor fuel geometry. Mo additions to UO2 have been shown to increase the thermal conductivity, thus reducing centerline temperatures and temperature gradients. Previous studies evaluated uniformly dispersed Mo or continuous Mo internal geometries (e.g., fins, plates, discs) that were selected using engineering intuition. The current study uses two different implementations of the same GA to optimize Mo placement and minimize the fuel temperature with the only constraint being a maximum 10% Mo volume fraction. One approach superimposed Mo line elements onto the monolithic UO2 pellet model, and the other converted entire UO2 volume elements to Mo. The former method generated 1D heat transfer connections between nodes, whereas the latter method allowed for the formation of 3D structures. Features of the optimal fuel design produced by the GAs included dispersed Mo near the centerline that shifted the peak fuel temperature outward by 0.6 mm, Mo chains in the high-heat-flux region in the mid-to-outer radial zone, and a large continuous structure that spanned the full radius and height of the pellet and accounted for 87.7 % of the total Mo in the pellet. Analysis of this design indicates that the optimal Mo configuration is a balance between creating continuous heat transfer pathways and optimally dispersing Mo to minimize the heat transfer distance through UO2. This architecture ultimately produced an effective thermal conductivity of 11.3 W/m·K under the assumed boundary conditions. This result is higher than any previous values from the literature. Potential fabrication methods and challenges are discussed in addition to the implications on fuel performance.