Operation Of Pressurized Water Reactor Research Articles

Development of a Zr-Nb-H-O reactive force field for molecular dynamics simulations of in-reactor corrosion

The corrosion of Zircaloy fuel cladding is one of the crucial issues during the operation of pressurized water reactors (PWR). In this work, a Zr-Nb-H-O reactive force field (ReaxFF) is constructed to study the corrosion mechanisms and irradiation effect of Zr-Nb alloys at the atomic landscape. This ReaxFF can describe the interactions between water dissociation products and Zr-Nb alloys, and the stability as well as diffusion properties of irradiation defects in Zr bulk. The molecular dynamics (MD) simulations based on the present ReaxFF show that Nb can thicken the suboxide layer during the corrosion process, as well as the promotion effect of irradiation on corrosion varies under different PKA (Primary Knock-on Atom) energies. The ReaxFF described here has the potential to be a new tool for understanding the in-reactor corrosion mechanism under the irradiation environment.

The current status of zirconium alloys fuel cladding

Properties such as corrosion and radiation resistance of zirconium alloys fuel claddings for nuclear reactors and ways to improve them are considered. The need to ensure high radiation and corrosion resistance and reliability of fuel claddings for the safe operation of pressurized water reactors of NPPs under operating conditions is substantiated. The use of sponge zirconium as the base of the alloys provides safety criteria under LOCA conditions. Optimization of the Fe content in the Zr-1%Nb alloy leads to increased corrosion resistance in the coolant and resistance to radiation growth during irradiation.

Experimental study on flow boiling CHF in annulus channel under heaving conditions using simulant fluid R134a targeting nuclear reactor applications

Floating nuclear power plants have attracted significant attention as a prospective means to deliver clean energy, particularly in geographically remote areas where conventional power infrastructure is inaccessible. The ocean environment, however, can affect the thermal–hydraulic phenomena in them. Notably, in the two-phase system, Critical Heat Flux (CHF) can change due to the periodic variation of additional forces. While a few experimental studies have been conducted on CHF under heaving motion, expanding the range of test conditions is necessary to elucidate its mechanism at the operating pressure of a Pressurized Water Reactor (PWR). Hence, the present study performed an experiment on the flow boiling CHF under heaving motion. To simulate the heaving motion of the ocean environment, a heaving motion platform was designed and constructed. It can simulate the heaving motion with a maximum acceleration of 0.6 g and a period ranging from 3 to 6 s. The CHF test loop utilized refrigerant R134a as the working fluid, and the tests were conducted under the thermal–hydraulic conditions corresponding to PWR operation conditions applying a fluid-to-fluid scaling. The test section consists of a single heater rod in the annulus channel. CHF was measured under both static and heaving motion conditions, and a comprehensive parametric study was conducted on the CHF variation due to the oscillating acceleration field. The results demonstrated that, in most cases, CHF decreased compared to the static condition. The reduction in CHF due to heaving motion could reach up to 9% in the present test configuration, and the reduction ratio was proportional to the magnitude of the heaving acceleration. In addition, the effect of heaving motion was pronounced in two distinct thermal–hydraulic regions: when the critical quality approached zero and when it exceeded 0.6. Based on the experimental observations, the CHF reduction mechanism in each region was suggested.

Parallels between natural Oklo reaction zones and industrial reactors dynamics.

The first man-made nuclear reactor was developed by Fermi and collaborators at the University of Chicago and reached criticality in December 1942. This was the confirmation that men were able to use sustained fission reactions in order to produce energy. Following this success, nuclear reactors studies gave rise to several families of reactors corresponding to different orientations and technical choices. They are linked mainly to the choice of fuel (natural uranium, enriched uranium, plutonium, thorium), coolant (water, carbon dioxide, helium, sodium, ...) and moderator for slow neutron reactors (graphite, light water, heavy water). Out of all these choices, the pressurized water reactor (PWR) family is the closest to the Oklo natural reactors. Many intriguing similarities are observed and discussed in the present article. Our present-day understanding of the PWR operating conditions has been a great help for understanding the Oklo reactors.

A novel method for calculating annular fuel cell resonance effective temperature

Annular fuels can increase the power density of reactors because of their double-sided cooling design, which is of great significance to the miniaturization and long-life operation of pressurized water reactors. In this study, a thermal–hydraulic analysis code named Thermal-Hydraulic Code of Annular Fuel with Single channel (THCAFS) was developed for annular fuels, and a calculation model for the resonance effective temperature of annular fuel cells was established. An annular fuel cell of a four-loop pressurized water reactor (PWR) designed by Westinghouse company was used as the calculation object, and the numerical results of THCAFS were compared with those of existing sub-channel codes such as VIPRE-01, TAFAX, and NACAF. Meanwhile, the Monte Carlo code RMC was used to simulate the radial power distribution and burnup process of the annular fuel, and its thermodynamic behavior was simulated using the self-developed code THCAFS. The radial power distribution, nuclide density variation, and cell temperature field under different burnup levels were also obtained, and the reaction rate equivalence was used instead of the effective resonance integral to investigate the calculation method used for resonance effective temperature of the annular fuel cell. The results show that THCAFS can be preliminarily applied to annular fuel designs and thermal–hydraulic analysis, while the effective temperature calculation method can provide important reference values for studying the resonance effective temperature of annular fuels.

The DISNY facility for sub-cooled flow boiling performance analysis of CRUD deposited zirconium alloy cladding under pressurized water reactor condition: Design, construction, and operation

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

Discovering the causes for the change of the vibration characteristics of the core support barrel in PWR nuclear power plants: A combined investigation based on ex-core neutron noise analysis and numerical modal analysis

As a key component of pressurized water reactor (PWR) nuclear power plants, the motion of the core support barrel (CSB) has a great impact on the safety of the nuclear reactor. However, the vibration characteristics of the CSB have not yet been fully investigated and may change due to the high doses of radiation and high temperature in the reactor. In this paper, the ex-core neutron noise signals of two PWR units throughout the multiple fuel cycles were analyzed and the vibration frequency of the CSB beam mode was found, for the first time, to drop with time during the operation of PWRs. Then, the statics analysis of the CSB was conducted to preliminarily identify the potential causes. For the further clarification of such vibration frequency drop phenomenon, a numerical assembly model of PWR reactor internals, including the CSB, hold-down ring (HDR), fuel assembly hold-down spring (FAHDS), etc., was proposed with consideration of the frictional contact and fluid-structure interaction for the modal analysis of the CSB. The results showed that the decrease of the FAHDS stiffness is the main cause for the drop of the vibration frequency of the CSB beam mode in real conditions.

Prediction of inter-granular crack initiation susceptibility in post irradiated grade 300 steels

Grade 300 austenitic steel components are subjected to irradiation-assisted stress corrosion cracking hazard, in the context of nuclear pressurized water reactor operation. In this paper, it is assumed that inter-granular cracking susceptibility primarily depends on sub grain plasticity mechanisms, controlling the slip band thickness and spacing achieved for a plastic strain level imposed. A computational model evaluating the related grain-boundary stress evolutions is developed and evaluated, using actual post-irradiation observations. In practice, we calculate a quantitative damage susceptibility factor R depending on the characteristic local slip band arrangements and the grain to grain misorientation. The statistical R evolutions obtained are consistent with the strain localization and related intergranular crack susceptibility trends, under corresponding irradiation and deformation conditions.

Estimation of tritium generation and discharge of the AP1000 reactor based on historical discharge data from the U.S. Pressurized water reactors

During the normal operation of pressurized water reactors, tritium has contributed more than 95 % of the total radioactivity of all the radionuclides discharged into the environment and has been recognized as the key radionuclide in the design and operation of reactors. In the paper, the tritium production of the AP1000 reactor by the Westinghouse approach has been introduced and the diffusion fractions of tritium from fuel rods in the approach have been reviewed according to advances in research on the diffusion of tritium from zirconium alloys. The historical tritium discharge data from the predecessors with very similar core designs of the AP1000 reactor have been collected and sorted. According to the statistical results, the average tritium discharge approximates the tritium generated from neutron activation of boric acid and lithium hydroxide and it shows that the release fraction of tritium through fuel cladding with zirconium alloy may be neglected for the best-estimated condition. The release of tritium from secondary neutron sources has been validated by a comparison of historical tritium discharge from the predecessors with and without secondary sources. In addition, it indicates that local maximum tritium emissions from the U.S. Pressurized Water Reactors are remarkably affected by batch liquid release, especially before or during the overhauls. It will help recognize the tritium generation in the primary loops and optimize the management of radioactive emissions for the utilities of the AP1000 reactor in the future.

파울링된 피복관 표면의 가압 경수로 운전환경에서의 미포화 비등 열전달 성능평가를 위한 실험장치 구축 연구

The CRUD (Chalk River Unidentified Deposit or Corrosion Related Unidentified Deposit) is a fouling deposit of metal or metal oxide on the nuclear fuel cladding under PWR (Pressurized Water Reactor) operating conditions and causing the several operational and safety-related problems. The major concerns related to the CRUD are AOA (Axial Offset Anomaly) caused by the accumulation of the boron species inside the porous CRUD from the primary reactor coolant and CILC (CRUD Induced Localized Corrosion) caused by locally increased cladding surface temperature due to the additional thermal resistance of the CRUD layer. The growth of CRUD and the CRUD-related problems are known to be governed by the sub-cooled nucleate boiling phenomena inside the upper part of the reactor core, especially on the hot fuel assembly and hot fuel pin. In the current study, the experimental facility called DISNY (crud DeposItion Simulator for Nuclear energY) was proposed as a testbed to simulate the CRUD growth and to investigate the sub-cooled nucleate boiling heat transfer performance of the CRUD deposited cladding surface under prototypical PWR operating conditions.

Study on water level control system of natural circulation steam generator

The natural circulation U-tube steam generator (UTSG) plays an important role in the operation and safety of pressurized water reactor (PWR) nuclear power plants (NPPs). It is very difficult to control the water level of UTSG during the operation of PWR. Therefore, it is necessary to study its dynamic characteristics. Based on the mass, energy and momentum conservation equations, the nonlinear model (NM) and transfer function model (TM) of UTSG are obtained through mechanism derivation. By introducing typical disturbance into the TM, the changing trend of transient response conforms to the physical law, and the dynamic characteristics are obtained based on the transient results. Based on time domain and frequency domain analysis, the water level control system (WLCS) of UTSG is designed, including the control scheme and controller parameters. The simulation results show that the performance of the WLCS is better than the traditional one. The WLCS not only solves the problem of poor linearity of the feedwater regulating valve (FRV), but also solves the problem of controller redesigning after the FRV replacing. The control system can be switched automatically at 15% power level.

Design of an Output Feedback Pressurizer Controller With a Decoupling Precompensator for PWRs

The pressurizer is a crucial component for the safe operation of pressurized water reactors (PWRs). The pressure and level control systems of the pressurizer interact strongly with each other. This cross-coupling between the inputs and outputs makes it difficult for conventional control methods to achieve optimal control effects on both the pressure and level. A simple state feedback control for the pressurizer with a decoupling pressure level precompensator realized using the diagonal matrix method is therefore proposed in this study. A detailed three-zone nonequilibrium dynamic model that reflects the stratification and mass and energy exchange phenomena inside the pressurizer is applied, and a state feedback control that utilizes the physically measurable states of the pressurizer is employed to simplify the state feedback controller to a form that is equivalent to the output feedback control. The improvement of the existing dominant pole placement technique in the feedback controller is described, and the correlation between the parameters and automated tuning methods is also given. The performance of the control system is evaluated through simulations under different operating scenarios to verify the proposed method.

Design and comparison of two advanced core control systems for flexible operation of pressurized water reactors

This paper focuses on the design of advanced core control systems for future generations of pressurized water reactors. The objective is to improve the flexibility of nuclear power plants to cope with the rapid growth of renewable energies. In practice, this means that the average coolant temperature, the axial power distribution of the reactor core and the position of the control rods have to be properly regulated during power variations. In previous work, conducted by the same authors, two promising approaches were investigated: 1) fixed-structure gain-scheduled control and 2) nonlinear model predictive control. Here, both methods are tested according to industry standards in an attempt to determine the best one for our problem. To achieve this, two different controllers are designed using a new multipoint kinetic model of the reactor core, which provides an accurate representation of the axial power distribution. The advantages and drawbacks of both design methodologies are discussed and then compared on PWRSimu, an intermediate complexity pressurized water reactor simulator developed by Framatome.

Design and application of supervisory control based on neural network PID controllers for pressurizer system

The pressurizer system (PRZ) plays a vital role in the operation of pressurized water reactors (PWRs). The PRZ is inherently a non-linear, time-variant system, and hence the PRZ modeling with transfer function or linear time-invariant state-space forms is not accurate enough. In this paper, the thermal-hydraulic behavior of the PRZ is modeled by RELAP5 thermal-hydraulic code that is one of the best modeling codes offered for the PRZ system. Although the RELAP5 best-estimate code is appropriate for PRZ modeling, the RELAP5 is not capable to implement the advanced controllers. As first-time, to eliminate this issue, the RELAP5 code and MATLAB software are coupled to use the capabilities of MATLAB (the ability to implement advanced controllers) and RELAP5 (the best model for the PRZ system) simultaneously. Accordingly, this coupling provides a new platform for designing and implementing various intelligent and advanced controllers for PRZ pressure and level in RELAP5 code. Likewise, after providing the platform for implementing intelligent controllers, a novel supervisory control based on neural network PID (NN-PID) controllers is designed for the PRZ pressure and level control. Also, the PRZ system is modeled as a MIMO system with consideration of the PRZ pressure and level interactions. Furthermore, as a case study, the advanced designed control system is applied to adjust the pressure and level of PRZ in a VVER-1000 type nuclear power plant. The study results show that this newly designed control system is able to control the PRZ pressure and level efficiently and effectively in several conditions. Therefore, the proposed methodology as a novel design of PRZ control can be applied and developed in new generations of nuclear power plants and pressurized test loops.

A data-driven strategy for xenon dynamical forecasting using dynamic mode decomposition

Detection and prediction of xenon induced oscillations are an important part in the operation of pressurized water reactors. Several models have been proposed for the prediction or estimation of xenon oscillations with drawbacks e.g. strongly depend on the initial xenon and iodine distributions, hard-to-implement or computationally costly. In this article, we proposed a fast, model-free and easy-to-implement data-driven strategy based on dynamic mode decomposition (DMD) to forecast the power distribution during the process of xenon oscillations. Various comparative experiments based on HPR1000 reactor show that the proposed data driven strategy is able to capture the complex relational characteristics of the temporal and spatial data of xenon oscillations. It can efficiently discover the hidden dynamicities and thus offers an accurate prediction of the system behavior.

SMART with Trans-Critical CO2 power conversion system for maritime propulsion in Northern Sea Route, part 1: System design

The Northern Sea Route is a new way to connect Asia and Europe. To utilize the sea route, merchant ships are required to have the icebreaking capability. Since conventional fossil fuel engines have limitations for the icebreaking merchant ship propulsion, nuclear power having a long refueling period for icebreaking merchant ship is proposed in this paper. To evaluate the technical feasibility of the nuclear power for the icebreaking ship’s engine, 330MWth SMART is selected for the reference reactor core. The power conversion system for the SMART core is substituted with the trans-critical CO2 recompression cycle that has advantageous in low-temperature heat sink as in Northern Sea Route and low-temperature heat source, Pressurized Water Reactor operating conditions. In this paper, the major components of the trans-critical CO2 cycle are designed. Then, CO2 PRHRS, which is driven by CO2 natural circulation loop, is designed to substitute steam-water based PRHRS.

A New Repair Criterion for Inconel690 Alloy Steam Generator Tube with Fretting Wear

Structural integrity assessment of steam generator tubes(SGT) is important for the safety operation of pressurized water reactor. In this work, a repair criterion for Inconel690 SGT with fretting flaw was proposed. Firstly, experiments were conducted to address material strength, eddy current examination and bursting pressure of defected SGT. Then, a numerical method was proposed to predict bursting pressure. At last, all factors were concluded in a simplified statistical method of structural integrity assessment. For anti-vibration bar fretting, Condition Monitoring criterion, Operational Assessment criterion as well as the repair criterion for different inspection interval were proposed. The present criterion can provide reference for the inspection interval considering long-term temporary shutdown or standby.

A functional variable universe fuzzy PID controller for load following operation of PWR with the multiple model

To achieve core power control of Pressurized Water Reactor (PWR) over full power range under arbitrary initial power conditions, a global controller under the multiple model for reactor core load following is designed. Based on the nonlinear model of PWR core, the transfer function models at five power levels are chosen as local models of reactor core. The multiple model of PWR core in the global range of power level is the weighted sum of five local models with the triangle membership functions. Based on the multiple model of PWR core, a global controller, is contrived by five local controllers at different core power level with the Functional Variable Universe Fuzzy PID (FVUF-PID) controller. The simulation results show that the global controller has good convergence and performance robustness characteristics over a wide range of operational parameters.

Efficient Linearized Model of Pressurizer System in Pressurized Water Reactors for Control Purposes

The pressurizer has an important role to ensure the protected operation of pressurized water reactor (PWR) by keeping the reactor coolant system pressure among allowed tolerances. In this paper, a non-equilibrium two-region nonlinear pressurizer (PZR) model was linearized using Taylor technique to introduce the linear models of pressurizer transients for the controller design purposes. Taylor series expansion of functions around an equilibrium point was calculated. The same assumptions of the reference model are considered in linearized model. The linearized model consists of five states; rate of changes for: mass of steam &water in the PZR, mass of water in the PZR, mass of water in the PC, steam temperature in PZR, and water temperature in PZR. The nonlinear and linear PZR models was implemented in MATLAB/Simulink. The linearized model of PZR is verified. The performance of linearized model of PZR and nonlinear model is consistent with small and acceptable errors. The closed-loop of the linearized model of PZR pressure and water level is implemented and investigated using standard Proportion-Integration-Differentiation (PID) controllers. Simulation results and the evaluation performance indicate that the proposed linear model of PZR is efficient, and it can be used for control purposes.

Optimal feed-water level control for steam generator in nuclear power plant based on meta-heuristic optimization

Optimal control systems are crucial requirement that emphasize the safe operation of pressurized water reactor (PWR) nuclear power plants. The present study focuses on steam generator (SG) feedwate...