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A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were

A study was performed comparing the performance of UO2 and (Th,U-233)O2 fuel in small long-life PWR. The neutronic calculation carried out by PIJ module of SRAC2006 was done to a fuel cell in 10 years of operation. The calculation was conducted by varying the enrichment of U-235 in UO2 and U-233 in (Th,U-233)O2 for 1% - 20% and also by varying the fuel volume fraction for 40%, 45%, 50%, 55%, and 60%. The performance was observed by comparing the enrichment needed by each fuel type to gain criticality in 10 years, the infinite multiplication factor (k-inf) value, and the conversion ratio (CR) value. The calculation results showed that 60% fuel volume fraction gave critical conditions with the lowest infinite multiplication factor and highest conversion ratio for both fuel types. While in terms of fissile nuclide enrichment needed, (Th,U-233)O2 had better performance than UO2, because only 5% U-233 was needed in (Th,U-233)O2 while UO2 needed 9% U-235 to gain criticality in 10 years of operation.

BackgroundSilicon carbide (SiC) composite claddings are candidate solutions for accident resistant fuel claddings in light water reactors.PurposeThis study aims to estimate the failure probability of a double-layer structured SiC cladding under a loss-of-coolant accident (LOCA).MethodsBased on a failure probability calculation method for SiC composite cladding, a quasi-steady state method was used to simulate and calculate the SiC composite cladding failure probability under transient conditions. Sensitivity analysis of the two characteristic parameters of Weibull distribution was performed by analyzing the proportion of various stresses under accident conditions. The effects of different burn-up conditions on the failure probability were investigated, and the failure probability of the cladding under different layer thickness ratios was simulated.Results & ConclusionsSimulation results indicate that the transient failure probability of SiC composite claddings is significantly affected by changes in the proportion of the composite layer and Weibull parameter, as well as the occurrence of LOCAs under different burn-up conditions. This study makes contribution to the development and design of accident resistant fuel claddings, providing reference for further investigations on the failure probability of SiC composite claddings.

Fouling due to magnetite deposition has been a major concern for steam generator (SG) tubing of pressurized water reactors (PWRs). Alloy 690 SG tubes are now used for new plants or are scheduled to replace old Alloy 600 tubes of operating plants. The purpose of this study is to investigate the magnetite deposition behavior on the two different SG tube materials: Alloy 600 and Alloy 690. Deposition tests were conducted under a sub-cooled nucleate flow boiling condition in simulated secondary water of a PWR at 270 °C. After these tests, we observed that the tube surfaces were covered with deposits composed of porous magnetite particles. We found approximately 30% more magnetite deposits on Alloy 600 than on Alloy 690. The electrostatic repulsive force between the magnetite particles and the Alloy 600 surface was only half of that between the magnetite particles and the Alloy 690 surface, resulting in an increase in the deposit mass. [ABSTRACT FROM AUTHOR]

Tracking load changes in a pressurized water reactor (PWR) with the help of an efficient core power control scheme in a nuclear power station is very important. The reason is that it is challenging to maintain a stable core power according to the reference value within an acceptable tolerance for the safety of PWR. To overcome the uncertainties, a non-integer-based fractional order control method is demonstrated to control the core power of PWR. The available dynamic model of the reactor core is used in this analysis. Core power is controlled using a modified state feedback approach with a non-integer integral scheme through two different approximations, CRONE (Commande Robuste d'Ordre Non Entier, meaning Non-integer order Robust Control) and FOMCON (non-integer order modeling and control). Simulation results are produced using MATLAB® program. Both non-integer results are compared with an integer order PI (Proportional Integral) algorithm to justify the effectiveness of the proposed scheme. Sate-space model Core power control Non-integer control Pressurized water reactor PI controller CRONE FOMCON. [ABSTRACT FROM AUTHOR]

The creation of methods for automated power control of power units is an urgent task, for which it is advisable to use the capacities of nuclear power plants. A mathematical model of a nuclear power plant (NPP) as a control object is proposed, which includes a multi-zone model of the active zone with distributed parameters, which makes it possible to take into account its internal properties (including transitional processes for xenon). This makes it possible to reduce the error in modeling the static and dynamic properties of nuclear power plants. A method for automated control of NPP power change using three control loops has been developed: one maintains a scheduled change in reactor power by controlling the concentration of boric acid in the coolant, the second maintains the required value of the axial offset by changing the position of the adjustment rods, and the third supports the temperature regime of heat transfer. Due to the adjustment of the position of the main valves of the turbogenerator, the developed method makes it possible to improve the stability of the energy release in the core with a change in its power under normal operating conditions of the reactor. [ABSTRACT FROM AUTHOR]

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency.Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

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]

The article is devoted to the determination of physiochemical forms of iodine radionuclides in the primary loop coolant of transportable nuclear power plants (TNPP) of pressurized water type with ammonia water chemistry. The main form of iodine radionuclides 131I–135I in the coolant is the iodide form. The content of iodate ions is 1–2% at an ammonia concentration of 18 to 79 mg/kg. The fraction of volatile organic and inorganic forms of iodine radionuclides passing into the gas phase from the coolant does not exceed 3.3 × 10–2 and 2.5 × 10–4%, respectively. [ABSTRACT FROM AUTHOR]

In pressurized water reactor (PWR) system, the surgeline plays an important role in bonding the pressurizer and the primary circle. Some considerable problems, including the thermo-hydraulics, the thermal stratification and the accompanying thermal stress under transient conditions, pose risks to the surgeline integrity. Herein, a fully-coupled flow-heat-thermo-elasticity model was developed to investigate the transient behavior of thermo-hydraulic parameters and the thermal stratification phenomenon in PWR. To evaluate the nonuniformity of the stratified flow, a stratification degree indicator ζ is introduced. It is found that during the outsurge flow, the increase of temperature variation will enlarge the temperature gradient on the wall, corresponding to a more serious deformation. In the cases of positive temperature variation, the higher temperature variation causes higher stratification degree ζ, and vice versa. The mass flow rate ṁ and the stratification degree are in negative correlation. The local deformation can reach 1.802 cm under a 50 K temperature variation, while its location varies from case to case. More attention should be paid to the regulation between the highest deformation location and the surgeline thermo-hydraulic parameters. [ABSTRACT FROM AUTHOR]

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

The microstructure and texture of three 316L foils of 25 μm thickness, which were subjected to different manufacturing process, were systematically characterized using advance analytical techniques. Then, the electrochemical property of the 316L foils in simulated pressurized water reactor (PWR) solution was analyzed using potentiodynamic polarization. The results showed that final rolling strain and annealing temperature had evident effect on grain size, fraction of recrystallization, grain boundary type and texture distribution. It was suggested that large final rolling strain could transfer Brass texture to Copper texture; low annealing temperature could limit the formation of preferable orientations in the rolling process to reduce anisotropy. Potentiodynamic polarization test showed that all samples exhibited good corrosion performance in the simulated primary PWR solution.

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design- basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCMfuel in a 17 X 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to- diameter ratio (P/D) and downselected two alternate assembly designs: 11 X 11 (1.65-cm outer diameter and 1.18 P/D) and 9 X 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

Radioactive corrosion product materials collected from the control rod drive mechanism (CRDM) housing in a pressurized water reactor (PWR, HANBIT-1 KNPP) were analyzed using an electron probe micro analyzer (EPMA). It is challenging to analyze the composition of radioactive corrosion products using an EPMA due to the rough surface shape and size, and even more so when the products are stacked in the form of small grains. The purpose of this study is to determine whether the corrosive products found inside the CRDM housing are stuck in contact with primary coolant or just oxide. In this study, not only was the surface condition of the samples very rough, but the samples that were quantitatively analyzed using a normal method had extremely low electrical conductivity using a normal method. We therefore tested a new semiquantitative analysis method using X-ray image mapping. In this technical note, we propose a method for collecting and analyzing corrosion products adsorbed in the CRDM. Reference papers on radioactive corrosion products collected from the CRDM could not be found. It is consequently difficult to argue that the method of collecting samples and performing the quantitative analysis suggested in this study is the best, but it can be said that it is an appropriate analysis method. Finally, the usefulness of the semiquantitative analysis is reviewed by verifying the analysis results of radioactive corrosion products collected from the CRDM housing in a PWR. [ABSTRACT FROM AUTHOR]

Applying an accurate parametric prediction model to identify abnormal or false pressurizer water levels (PWLs) is critical to the safe operation of marine pressurized water reactors (PWRs). Recently, deep-learning-based models have proved to be a powerful feature extractor to perform high-accuracy prediction. However, the effectiveness of models still suffers from two issues in PWL prediction: the correlations shifting over time between PWL and other feature parameters, and the example imbalance between fluctuation examples (minority) and stable examples (majority). To address these problems, we propose a cost-sensitive mechanism to facilitate the model to learn the feature representation of later examples and fluctuation examples. By weighting the standard mean square error loss with a cost-sensitive factor, we develop a Cost-Sensitive Long Short-Term Memory (CSLSTM) model to predict the PWL of PWRs. The overall performance of the CSLSTM is assessed by a variety of evaluation metrics with the experimental data collected from a marine PWR simulator. The comparisons with the Long Short-Term Memory (LSTM) model and the Support Vector Regression (SVR) model demonstrate the effectiveness of the CSLSTM.

BackgroundUnder the condition of sub-cooled nucleate boiling (SNB), corrosion products in primary coolant of nuclear reactor will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Previous literature shown, zinc injection in primary coolant is an important method to inhibit the fuel crud deposition on the fuel cladding surface.PurposeThis study aims to investigate the influence of zinc concentration on the behavior of fuel crud deposition, and eventually provide guidance for zinc injection in primary coolant of nuclear power plant.MethodsThe fuel crud deposition tests of domestic zirconium alloy fuel cladding in different zinc concentrations were carried out by using a self-made fuel crud deposition device. Tubular crud deposition test specimen with built-in heating unit was designed and prepared for simulation study. After the tests, stereo microscope (SM) and scanning electron microscope (SEM) were employed to observe the macro and micro morphology of fuel crud whilst the composition of of fuel crud was observed and analyzed by the energy dispersive spectroscopy (EDS) with SEM, and X-ray photoelectron spectroscopy (XPS) was used to analyze the contents of Zn and B elements in the crud phase and inside the crud.ResultsObservation results show that the chimney-like crud formed on the fuel cladding surface becomes less obvious with increasing the zinc concentration in the coolant and the crud surface becomes flatter. Simutabeously, the crud thickness, the ratio of Ni/Fe and the boron precipitation mass within the crud are decreasing with increase of the zinc concentration. When the zinc concentration increases to 100 μg‧L-1, new Zn-containing phases precipitate within the crud.ConclusionsWithin the zinc concentration of 0~100 μg‧L-1, zinc injection in primary coolant of reactor can significantly inhibit the crud deposition on the fuel cladding surface.

BackgroundCompared with the traditional pressurized water reactor (PWR), the core design of large advanced PWR CAP1400 has significant changes, such as the increase in the number of fuel assemblies, the increase in reactor power, the increase in the average temperature of core coolant, etc. These changes have an important impact on the results of rod ejection accident, and then affect the safety of reactor.PurposeThis study aims to verify the safety of large advanced PWR under rod ejection accident condition and the influence of key input parameters on accident analysis results.MethodsBased on the neutron dynamics software TWINKLE and fuel performance analysis program FACTRAN, the typical four types of operating conditions, including the beginning of life the hot full power and the hot zero power, the end of life the hot full power and the hot zero power, were selected to carry out the simulation calculation of the control rod ejection accident analysis for large advanced PWR, and the sensitivity analysis of key input parameters of rod ejection accident conditions was performed by using the direct numerical perturbation method.Results & ConclusionsSimulation results show that the power peak is the most sensitive to the worth of rod ejection, but less sensitive to shutdown reactivity. The consequences of the control rod ejection accident designed for CAP1400 can meet the requirements of acceptance criteria and the reactor is in the safe and controllable state.

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The direct current potential drop (DCPD) technique may be used in crack propagation tests to measure the crack growth rate (CGR). Potential probes attached to the specimen allow the variation of the crack length to be estimated. In this research, the DCPD technique using one single potential probe was applied to solid bar specimens (i.e., without any initial notch or crack) subjected to low-cycle fatigue testing in a simulated pressurized water reactor (PWR) environment. This particular analysis had two associated difficulties, the first one being the fact that crack initiation sites are not known beforehand, and the second one consisting in the experimental difficulties and conditioning factors associated with the simulation of the PWR environment. Nine solid bar specimens were tested to fatigue failure under different strain amplitudes and frequencies, while also measuring the corresponding DCPD signal during the fatigue process. It was observed that the initiation of multiple cracks was detected by the DCPD measurements. Moreover, as fatigue continued, one of the cracks became dominant and progressed to cause the specimen failure. The DCPD technique allowed the average CGR of the dominant crack to be estimated. Finally, the obtained average CGRs were validated by comparing them with average CGRs derived from striation spacing measurements, obtained from scanning electron microscopy (SEM) and from literature values gathered in the NUREG/CR-6909 document.

The aim of this work is to introduce the novel concept of an m-polar fuzzy soft set, including various types of algorithms and their fundamental operations. We created mathematical modeling to analyze operational rules and discuss the advantages, disadvantages, and natural aspects of algorithms for six types of nuclear power plants. It has been determined that emerging trends and the benefits of algorithms are increasing step by step. The suggested modeling with an m-polar fuzzy soft set is integrated into the fuzzy mean environment to analyze the effect of the correlation between decision factors and decision results without an excessive duty cycle, thus minimizing energy use and other adverse effects. Based on a new group decision-making technique considering an asymmetric weight vector, we proved that Gas Cooled, Graphite-Moderated, and Pressurized Water Reactors are the optimal choices for nuclear power plants. In the end, a numerical illustration is provided for selecting the best photo to demonstrate the use of the generated technique and to exhibit its adequacy.

This paper proposes a fiber optic Fabry-Perot (F-P) strain sensing system using non-scan correlation demodulation applied to the health monitoring of the pressurized water reactor’s fuel assembly structures. The structural design and sensing mechanism analysis of the sensor were carried out, and the strain transfer model from the fuel sheet to the strain gauge was established. After the sensor fabrication and installation, the static tests have been conducted, and the results indicate that the sensing system can accurately measure the microstrain with a sensitivity of up to 12.6 nm/με at a high temperature (300 °C). The dynamic testing shows that the sensing system has a good frequency adaptation at 10–500 Hz. Thermal-hydraulic experiments show that the sensing system can run stably in a nuclear reactor, with high temperature, high pressure, and high-velocity flow flushing; additionally, the consistency deviation of the measured data is less than 1.5%.

• 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]

BackgroundFor a pressurized water reactor (PWR), under the synthesis effects of primary water chemistry, reactor operation mode, system facilities replacement, fuel design and management, the corrosion products in primary coolant will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Fuel crud formed on the cladding surface may result in crud induced localized corrosion (CILC), crud induced power shift (CIPS) and raise the primary source term. The deposition behavior of fuel crud is affected by both the corrosion products concentration and the core boiling rate.PurposeThis study aims to investigate the influences of core power variation on the corrosion products concentration and the core boiling rate, and eventually evaluate the fuel crud deposition states under different core powers.MethodsThe core crud mass and the maximum crud thickness on fuel cladding surface were estimated by the CAMPSIS, an independent crud behavior analysis software developed by China Nuclear Power Technology Research Institute Co., Ltd (CNPRI). Influence of core power variation and the combined effect of core boiling rate and corrosion product concentration were analyzed.ResultsComputational results that, within the range of 75%⁓100% full power (FP), the lower the core power is, the smaller the core crud mass will be, but the maximum crud thickness will increase. When the core power decreases to 50% FP, the maximum boiling rate is close to zero, and both of the core crud mass and the maximum crud thickness are minimum.ConclusionsThe core power reduction of a nuclear reactor may lead to the increase of local fuel crud thickness, and adversely affect the heat transfer between the fuel cladding and the coolant. It is suggested that the influence of core power reduction on fuel crud deposition should be considered if a nuclear reactor tends to operate at a lower core power.

Summary: The design of burnable poisons (BPs) can compensate for excess reactivity at the beginning of lifetime of nuclear reactors and flatten power distribution, which is especially important for long‐cycle nuclear reactors. The design of BPs requires the optimization of material type, purity, layout, axial division of the BPs, so that the reactivity introduction at the beginning of life, poisons residues at the end of life, fuel utilization and power flattening can be comprehensively optimized, which is a multi‐input multiobjective optimization problem. At present, the traditional optimization design mainly relies on the subjective experience and judgment of designers, which is complicated and time‐consuming. Therefore, the efficiency and reliability of BPs design urgently need to be improved. In this article, the mathematical model of multiobjective optimization based on genetic algorithm (GA) was established for BPs design in pressurized water reactor (PWR) fuel assembly. Then, optimization program was developed by combining parallel multiobjective GA with Monte Carlo particle transport code Reactor Monte Carlo as the neutronics and depletion solver. The optimization method and program were applied to the BPs design in two‐dimensional and three‐dimensional fuel assembly of PWR. The optimization schemes of BPs searched by GA were similar to the schemes by the manual search of designers for two‐dimensional assembly in the previous research. The developed optimization method and program were proved to be effective for BP designs of PWR assembly, which do not require the manual experience of designers for searching the optimization scheme. This article provided useful methods and tools for BPs design of nuclear reactors. [ABSTRACT FROM AUTHOR]

Traditional light water reactor simulations are usually either high fidelity, requiring hundreds of node-hours, or low fidelity, requiring only seconds to run on a common workstation. In current research, it is desirable to combine the positive aspects of both of these simulation types while minimizing their associated negative costs. Because neural networks have shown significant success when applied to other fields, they could provide a means for combining these two classes of simulation. This paper describes a methodology for designing and training neural networks to predict normalized pin powers and k e f f within a reflective two-dimensional pressurized water reactor assembly model. The developed methodology combines computer vision approaches, modular neural network approaches, and hyperparameter optimization methods to intelligently design novel network architectures. This methodology has been used to develop a novel new architecture, LatticeNet, which is capable of predicting pin-resolved powers and k e f f at a high level of detail. The results produced by this novel architecture show the successful prediction of the target neutronics parameters under a variety of typical neutronics conditions, and they indicate a potential path forward for neural network–based model development. [ABSTRACT FROM AUTHOR]

This paper proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed technique employs a multivariant analysis technique, namely, principal component analysis (PCA) and clustering analysis (CA) to identify the position of the PWR control rod using its impact on the core radial thermal neutron flux along the axial track of motion. The results of these investigations have shown that the proposed technique successfully removed the limitation on the data size and any limitations imposed by outlier samples, extracted the noise, and provided near-instantaneous analytical and visual ways for position identification process with excellent generalization fitting and prediction efficiencies. In the context of this paper, multiple in-depth simulations are conducted to ascertain the efficiency of the proposed technique in identifying the control rod positions. These simulations have been conducted using a Westinghouse 2772-MW(thermal) PWR benchmark at 100% thermal power generation, where a three-dimensional TRITON FORTRAN-code has been utilized to simulate the radial thermal neutron flux of the PWR core. The PCA model is developed, tested, and generalized using the SIMCA software package. In addition, CA is also performed via the Minitab statistics software package in order to confirm the efficiency of the proposed technique. [ABSTRACT FROM AUTHOR]

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This paper is focused on overcoming reactor shutdown malfunction and diagnostics due to poor cooling and water drop level within a pressurized water reactor. So the temperature coupling analysis between the heat exchanger (HEX) and the U-tube steam generator (UTSG) is inspected under changes of primary and secondary water temperature. This coupling allows the removal of heat from the UTSG via the HEX. For the UTSG, implicit and explicit solutions for analysis and evaluation of UTSG characteristics are conducted. Scarce explicit models describing the behaviors of the UTSG are available. This analytical framework is proposed to control the water/steam flow within the UTSG of a nuclear power reactor (NPR). Exact performances for temperatures of metal tube temperature and primary water entering and exiting from the tube are derived. There is no one implicit model that can describe the performance characteristics of heat transfer within the UTSG. So a novel simulator declaring the operational behavior of the UTSG in an NPR has been built. This simulator of the UTSG provides exact handling of the UTSG performance characteristics. For the HEX, the exact handling of the HEX including boundary temperature and the primary and secondary interface temperatures are proposed. Three different models are implemented (combinational, steady state, and integral). The possibility of realizing higher steam quality is established through block diagram programming models. Compared to literature results, the built models are validated with high agreement. As a final conclusion, the proposed analyses allow the control of steam generation and flow. The introduced results compensate for the necessity of expensive and complicated controllers within the HEX and UTSG through parameter variation. Accordingly, the performance of the NPR is enhanced. [ABSTRACT FROM AUTHOR]

Best estimate plus uncertainty (BEPU) analysis method has been widely used to analyze various transient accidents of pressurized water reactor (PWR). However, the traditional BEPU method has some limitations: (1) The input parameters are not clearly defined, resulting in inaccurate conclusions in the sensitivity analysis. (2) The uncertainty quantification and sensitivity analysis usually share the same set of samples, but they have different requirements for the sample size. In this work, an improved BEPU method is proposed, which can alleviate the above defects. The improved BEPU method possesses the following two characteristics: (1) The sensitivity analysis is performed for the steady-state and transient calculation, respectively. It provides more comprehensive results than does the traditional BEPU method. (2) The sensitivity analysis is performed before the uncertainty analysis to reduce the number of uncertainty analysis inputs. A small-break loss-of-coolant accident (SBLOCA) is simulated by Reactor Excursion and Leak Analysis Program (RELAP) 5 to verify the accuracy and applicability of the improved BEPU method. By the sensitivity analysis, the coolant pump inlet roughness, main flow rate, core heat channel temperature, break area, and PRZ pressure have moderate or higher relationships with the peak core outlet temperature. The fission product yield factor has a moderate positive relationship with the peak cladding temperature (PCT). The sensitivity analysis by the improved BEPU method shows that the peak core outlet temperature has strong relationships with main flow rate, core heat channel temperature, and PRZ pressure, which is not captured by the traditional BEPU method. As a result, it is more reasonable to take steady-state parameters as inputs in the sensitivity analysis of transient. Only those parameters with high correlation coefficients are sampled for uncertainty analysis. Meanwhile, the results of the uncertainty analysis obtained by the improved BEPU are consistent with the results of the traditional method. Neither the PCT nor the peak core outlet temperature will exceed their limits. The results illustrate that the improved BEPU method can reduce the size of samples but maintains the desired accuracy.

A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.