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In this paper, energy, exergy, economic analysis and multiobjective optimization of a pressurized water reactor (PWR) nuclear cogeneration plant for hydrogen production through high-temperature steam electrolysis (HTSE) are carried out. HTSE requires energy in the form of both heat and electrical work. A novel parameter, namely the heat/total energy ratio, is defined, and used as a decision variable in optimization. In addition to energy ratio, hydrogen production capacity, reactor thermal power, live steam temperature, reheating mass flow rate ratio, reheating temperature and steam extraction location are considered as the decision variables to simultaneously optimize the thermal efficiency, thermal-to-hydrogen efficiency, utilization factor, exergy efficiency and total revenue of the cogeneration plant. The analysis and optimization focus on the secondary cycle of the PWR and the effects of hydrogen and electricity prices and ambient conditions are also taken into account since these prices have a significant impact on the optimum design. For a hydrogen price of 4 $/kg and an electricity price of 0.1 $/kWh, when equal preference is given to all objective functions, the optimum production capacity is 6.778 kg/s. The energy ratio has an optimum value if the optimization focuses exclusively on the thermal efficiency and total revenue. • Thermodynamic analysis and multiobjective optimization of a PWR nuclear hydrogen production plant are performed. • A novel parameter, heat/total energy ratio is defined for the high-temperature steam electrolysis process. • Increase in energy ratio improves thermal-to-hydrogen efficiency, exergy efficiency, utilization factor and total revenue. • For a hydrogen price of 4 $/kg and an electricity price of 0.1 $/kWh, the optimum production capacity is 6.778 kg/s. • Price independent optimization yields an optimum hydrogen production capacity range of 3.826–5.707 kg/s. [ABSTRACT FROM AUTHOR]

Obtaining real time quantitative condition assessment results is a focus in the field of nuclear power system PHM. The existing researches on the nuclear power system condition assessment based on the Multi-Criteria Decision-Making framework do not consider the faults perniciousness differences. This leads to the assessment results relying on the system status deviation degrees(severity), and the weighting and aggregation processes cannot effectively reflect the differences in the consequences and risks of different faults, which represent the faults harmfulness. A quantitative assessment method based on fault severity and fault harmfulness is proposed to address this issue. The novel method quantifies the fault severity based on the similarity principle while diagnosing system faults, and then analyzes the devices mainly affected by various faults to quantifies the harmfulness. Combining the system status deviation degrees(severity), the devices and parameters importance, and the differences in faults harmfulness, the novel method aggregates comprehensive system condition assessment results. Based on a widely recognized nuclear power system faults dataset, the novel method was compared with other methods. The conclusion is that the novel method considers the differences in the faults harmfulness, resulting in more reasonable assessment results and avoiding insufficient or excessive warnings. • An novel nuclear power system quantitative condition assessment method. • The weighting and aggregation consider the faults perniciousness differences. • Avoid insufficient or excessive warnings and support efficient maintenance. [ABSTRACT FROM AUTHOR]

• Higher residual stress/strain and dislocation density in alloy 600 HAZ than in base. • Thicker oxide and deeper intergranular oxidation in alloy 600 HAZ than in base. • Longer SCC length and higher cracking engagement in alloy 600 HAZ than in base. • Stress-assisted oxidation contributes to enhanced SCC growth in alloy 600 HAZ. • The impact of local mechanical parameters on SCC in the HAZ is highlighted. The impact of welding-induced mechanics on the oxidation and stress corrosion cracking (SCC) behavior of Alloy 600 in a pressurized water reactor (PWR) primary water was investigated using an Alloy 600-Alloy 152 M weldment. The microstructural analysis found a 0.05 mm-wide composition transition zone from welding. The Alloy 600 heat-affected zone (HAZ) grain size matches the Alloy 600 base (600B). Deformation hardening initially decreases and stabilizes from the fusion boundary (FB) to the 600B Comparing the microstructural characteristics of the specimen from the HAZ at about 0.5 mm from the FB with specimen 600B shows identical composition, but the former exhibits higher Vickers hardness (HV), kernel average misorientation (KAM), dislocation density, and residual stresses. Results from both short- and long-term oxidation tests indicate that the specimen from the HAZ exhibits a greater thickness of the inner oxide film, a higher number of local oxidation sites, and an increased maximum depth of intergranular (IG) oxidation compared to the 600B specimen. Multiple sets of SCC test results demonstrate that the crack growth rate (CGR) in the specimens from the HAZ is higher than that in the base, with the former showing longer IG cracks and a greater proportion of IG cracking. [ABSTRACT FROM AUTHOR]

In this study, the corrosion behavior of pre-ion-irradiated 304 SS is compared to that of un-irradiated 304 SS in simulated Pressurized Water Reactor (PWR) conditions for both early stages of corrosion and longer exposure times. Transmission electron microscopy is used to characterize the oxide scale. The study shows that although the pre-irradiated microstructure does not change the nature of the oxide and its dual-layer structure (i.e. inner/outer oxide layers) formed on 304 SS, the path of growth of the inner oxide layer is affected by the irradiated microstructure resulting in oxide protrusions into the matrix; a mechanism is proposed. • A dual oxide structure forms on 304 SS after corrosion in PWR conditions. • The dual layer shows a continuous Cr-rich inner oxide and Fe-rich oxide islands on the surface. • Pre-irradiation to 10 dpa does not alter the dual-layer oxide structure but affects the inner oxide thickness and shape. • Inner oxide protrusions into the matrix are observed below the bigger oxide particles. • Irradiation loops aligning under local stress and dislocation channels may aid oxygen diffusion. [ABSTRACT FROM AUTHOR]

This research investigates the performance of new radiografted materials as adsorbents for ionic and colloidal 110 mAg, a prominent pollutant in pressurized water reactor (PWR) wastewater. A comparative analysis was conducted to evaluate their effectiveness against commercially available resins. Due to the increasing demand for new nuclear power plants and the growing global dependence on nuclear energy, it is crucial to minimize radioactive waste production. The study was conducted on simulated inactive and radioactive waste samples with trace amounts of 110 mAg that are hard or impossible to detect by standard analytic techniques. The synthesized materials were labeled PGE (amine functional groups) and PGM (sulfhydryl and amine functional groups). Functionalization with amine and sulfhydryl groups was chosen since it is well known that these groups strongly interact with Ag in both ionic and colloidal form. The tested materials demonstrate exceptional performance as adsorbents by completely removing (100 %) both ionic (PGM) and colloidal (PGE) forms of 110 mAg. Additionally, depending on the adsorbed species excellent elution of 110 mAg was achieved (>94 % for ions and >50 % for colloids) implying possible material reusability. Manufacturing cost of such adsorbents is orders of magnitude lower per kg (10 – 50 $/kg), compared to prices of commercial resins (100 – 10000 $/kg). These results indicate that such inexpensive materials could potentially be used in an efficient and selective way to treat PWR wastewater for 110 mAg contamination while reducing the operating cost and minimizing resulting solid waste. [Display omitted] • Radiografted materials (PGM/PGE) allow complete 110 mAg removal from simulated nuclear wastewater samples. • Total adsorption of colloidal 110 mAg (100 %) is achieved, in contrast to standardly applied processing techniques (> 98 %). • Inexpensive production, robustness and potential reusability compared to commercial resins. • Reduction in solid waste production and complete mitigation of 110 mAg environmental contamination. [ABSTRACT FROM AUTHOR]

Simulated pressurized water reactor conditions (330 °C, 15.6 MPa, ∼20 ppb oxygen) without irradiation were used to investigate the hydrothermal corrosion behavior of ultrasonic additively manufactured Zircaloy-4 up to 1000 h. X-ray computed tomography allowed for visualization of defects from processing and their progression after corrosion experiments. The specimens were found to have clear variability in the mass change data, compared to typical wrought Zircaloy-4 specimens. The variation in the mass change after exposure was attributed to weld defects connected to the specimen surface which allowed ingress of oxidant into the samples. Defects visualized by computed tomography were found via metallography and characterized. Ultrasonic additively manufactured Zircaloy-4 was found to have comparable corrosion behavior as wrought Zircaloy-4 for specimens which did not have clear surface defects along weld interfaces. [ABSTRACT FROM AUTHOR]

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

Various Machine Learning (ML) techniques have seen recent and growing interest in the creation of surrogate neutronics models as a potential way to avoid the computational expenses associated with conventional high-fidelity modeling. Artificial Neural Networks (ANNs) have been shown to be particularly useful for single-assembly predictions involving pin-wise power distributions and multiplication factors. In this paper, the LatticeNet neural network framework is investigated as a method to predict Doppler and moderator temperature coefficients for Pressurized Water Reactor (PWR) fuel assemblies, as well as differential boron worth. This approach uses the built-in tools developed alongside LatticeNet to construct two fully-connected network architectures capable of predicting k -eigenvalues from two inputs such as fuel enrichment and temperature when trained with data generated with CASMO-4E. A single network taking in all study parameters as inputs was then used to predict k -eigenvalues for fuel temperature, moderator temperature, and boron perturbation cases. The calculated temperature coefficients and differential boron worth values were compared with a bank of reference values to validate the efficacy of this method. Overall, differences in k -eigenvalues were within 0.017% in the worst case. The temperature coefficients saw mean errors of 1.85% and 1.69% for the two-parameter networks, respectively. The all-parameter network was then shown to predict 1100 points in 236 ms compared to the 4.95 min CASMO-4E took to generate them. Additionally, the differential boron worth achieved the lowest mean error of 0.30%; each of these values were within our acceptance criteria. [ABSTRACT FROM AUTHOR]

• PWR water reduces the fatigue life by a factor of ∼10 compared to air environment. • Low strain rate and high SSR fraction increase FCGR and decrease fatigue life. • Aggregated DIMT in specimens tested in PWR water with short fatigue life. • H reduces SFE, increases ε-martensite density and enhances α'-martensite formation. • Martensite's susceptibility to H-assisted cracking enhances FCGR in PWR water. Austenitic stainless steels used in light water reactor coolant environments are susceptible to environmentally assisted fatigue due to non-monotonic loading conditions, primarily associated with load-follow, thermal transients, or intermittent plant shutdowns and start-ups. This study investigates the effects of a high-temperature pressurized water reactor (PWR) water environment and cyclic loading parameters on the low cycle fatigue behavior of austenitic 304L stainless steel. Prolonged exposure to a PWR environment and cyclic loading conditions such as a lower strain rate or a higher fraction of slow strain rate enhances the initiation and accelerates the crack growth rate of fatigue cracks, resulting in decreased fatigue life. The deformation-induced α'-martensite is observed in proximity to fatigue crack tips primarily in specimens tested in simulated PWR primary water, while cellular dislocation structures are more frequently observed near crack tips in specimens tested in high-temperature air. The deformation-induced martensitic transformation from γ-austenite to α'-martensite, occurring via the precursor ε-martensite phase, contributes to the accelerated fatigue crack growth rate in a PWR environment with hydrogen. [ABSTRACT FROM AUTHOR]

Natural circulating steam generators are extensively applied in pressurized water reactor units, where the water level control is a tough challenge. During transients, the water level may change drastically, and its reverse dynamic characteristics may even cause safety concerns. This article studies the dynamic effects of several crucial impact factors on the level, including their gains, time constants, and coupling relationships. Considering the impact and the inter connections of these factors, a dynamic distributed parameter mechanism model for real-time water level estimation is constructed. It consists of three sub-models: the close-loop control system model, the mass and heat transfer model, and the momentum transfer model. Solving these models, one may yield the flow rates of the inputs and outputs of the downcomer channel. Then, the level is determined employing the dynamic conservation for the feedwater chamber. Pseudo-online simulation studies are carried out with the measurements from an operating 1000 MW unit during the full load rejection test. • Impact factors affecting the steam generator water level are studied quantitively. • A dynamic distributed parameter model of steam generator water level is established. • The proposed model is validated using measured data during load rejection transients. • Comparison suggests the accuracy of water level simulation with a mean error of 0.13%. • The model is suitable for online application since each-step simulation takes only 7 ms. [ABSTRACT FROM AUTHOR]

• A GPU-accelerated solution for the neutron diffusion equation was proposed. • The solution is based on the fine mesh Galerkin Finite Element Method (GFEM). • The resulting program GFEM_GPU was applied to the 2D PWR benchmarks IAEA, BIBLIS, and ZION. • GFEM_GPU provides iterative solutions up to 50% faster than the CPU version. • This approach provides high-level implementation, faster prototyping, and possible application to Loading Pattern Optimization. The In-Core Fuel Management or Loading Pattern Optimization (LPO) is the problem of finding an optimal configuration of nuclear fuel assemblies in a reactor core. Over the years Optimization Metaheuristics based on populations have been used successfully for solving the LPO. However, for such methods, thousands of evaluations of candidate loading patterns with reactor physics codes are required. In addition, more precise calculations are also desirable, which in principle may conflict with faster evaluations. For coping with both criteria simultaneously, a code based on the Galerkin Finite Element Method (GFEM) for solving the neutron diffusion equation designed for Graphics Processing Unit (GPU) is proposed in the present work. The resulting program GFEM_GPU has been implemented in Python language with a library for GPU-accelerated computing. A comparison between GFEM_GPU and the CPU version of GFEM regarding three nuclear reactor benchmarks (IAEA-2D, BIBLIS-2D, and ZION-2D) is provided and the accuracy of the results is compared to the Nodal Expansion Method (NEM). The implementation of the GFEM solve-phase with GFEM_GPU provides iterative solutions up to 50% faster than the CPU version depending on the discretization of the problem. This approach presents several advantages, such as a high-level implementation, faster prototyping, and application to LPO problems, while keeping the high accuracy provided by fine mesh methods. [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]

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]

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 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]

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]

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]

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

The in-service brittle fracture of a structural component of a nuclear power plant (NPP) is a problem of major concern in engineering. During nuclear energy generation, the wall of the nuclear reactor pressure vessel (RPV) is exposed to a hydrogenating environment leading to a fracture phenomenon known as hydrogen embrittlement (HE). This in-service failure is ruled by hydrogen diffusion from the hydrogenating source (the inner side of the RPV) towards certain places inside the vessel where hydrogen is accumulated and microstructural damage is located. The diffusion process is highly influenced by the stress and plastic strain distributions. For achieving a realistic estimation of the hydrogen accumulation by diffusion, both the in-service thermal stress and the manufacturing induced residual stress and strain (due to tempering heat treatment) must be taken into account. In this paper, a numerical analysis of the hydrogen diffusion assisted by stress and strain is carried out to find out the hydrogen accumulation within the wall of a real pressurized water reactor (PWR) for diverse heat treatment conditions. Results reveal the key role of the in-service thermal stress which enhances the hydrogen diffusion through the constituents' materials of a PWR pressure vessel. [ABSTRACT FROM AUTHOR]

Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is coating a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Through considering the loading patterns, loading amount and self-shielding effect and so on, when the outer diameter of MA is 0.520 cm, the transmutation rate of MA is 38.5%, it is the optimal configuration, firstly transmutation rate and loading amount of MA with the pattern of MA in the water gap of burnable poison rods are larger than ones with the pattern of MA substitute for part burnable poison. Secondly the highest transmutation rate means the least loading amount of MA and the lowest transmutation rate means the most loading amount of MA. In order to keep reasonable transmutation rate and loading amount, when the outer diameter of MA is 0.520 cm, it would not reduce dramatically the amount of MA and the MA loading of this loading pattern is equivalent to the discharge amount of five pressurized water reactor, at the same time, the transmutation rate can keep reasonable level. [ABSTRACT FROM AUTHOR]

The effect of prior single tensile overloading on the stress corrosion cracking behavior of cold rolled 316L in a simulated PWR water environment at 310 °C was investigated. SCC growth retardation by overloading was observed in cold rolled 316L specimens in both the T-L and L-T orientations. The stretch zone observed on the fracture surfaces of the overloaded specimens affected SCC propagation. The compressive residual stress induced by overloading process reduced the effective driving force of SCC propagation. The negative dK/da effect ahead of the crack tip likely contributes to the retardation of SCC growth. The duration of overloading is dependent on water chemistry and the local stress conditions. [ABSTRACT FROM AUTHOR]

Human error is one of the leading factors in nuclear accidents. However, on-line monitoring may help operators to detect and locate the abnormality on time, and thus it has the benefit of reducing misoperation, improve the safety of nuclear power plants and guide the maintenance plan. This work presents a distributed on-line monitoring and fault detection methodology based on modeling and simulation of nuclear plant’s sub-units. This methodology utilizes quantitative calculation from the simulation model and combines the measurements gotten from instrumentation and control system to detect incipient faults. Under normal operation, the method allows real-time tracking to set dynamic thresholds. When a failure occurs, the corresponding parameters can be compared to detect the fault. Furthermore, the root cause, fault type and degree of the fault could be identified and calculated while considering compensation from the control system and trends of measurements. The performance of this method was evaluated and verified by applying it to the Reactor Coolant System of a PWR and comparing its result with the one obtained from Principal Component Analysis (PCA) method. The result of simulation analysis under steady operation and transients separately show the effectiveness and accuracy of this methodology when compared with PCA method; therefore, laying an essential foundation for the prediction of fault trends. [ABSTRACT FROM AUTHOR]

The paper focuses on the presentation and comparison of basic nodal and expanded multi-nodal models of the Pressurized Water Reactor (PWR) core, which includes neutron kinetics, heat transfer between fuel and coolant, and internal and external reactivity feedback processes. In the expanded multi-nodal model, the authors introduce a novel approach to the implementation of thermal power distribution phenomena into the multi-node model of reactor core. This implementation has the form of thermal power distribution coefficients which approximate the thermal power generation profile in the reactor. It is assumed in the model that the thermal power distribution is proportional to the axial distribution of neutron flux in the un-rodded and rodded reactor core regions, as a result of control rod bank movements. In the paper, the authors propose a methodology to calculate those power distribution coefficients, which bases on numerical solutions of the transformed diffusion equations for the un-rodded and rodded reactor regions, respectively. Introducing power distribution coefficients into the expanded multi-nodal model allows to achieve advanced capabilities that can be efficiently used in design and synthesis of more advanced and complex control algorithms for PWR reactor core, for instance in the field of reactor temperature distribution control. [ABSTRACT FROM AUTHOR]

The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters. [ABSTRACT FROM AUTHOR]