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In this paper, a simplified algorithm is implemented to develop a probabilistic fracture mechanics code using python programming language. This code is able to predict the service life of a pressurized water reactor pressure vessel of Nuclear Power Plant considering its beltline region depending upon the material properties and operational conditions. Considering the manufacturing defect which is responsible for the initial flaws of a material, this code is able to calculate the failure probability which in turns gives the suggestion of the improvement in the manufacturing process. Besides, this PFM code which is available in different countries with their own version will help to estimate the safety margins of an operating pressure vessel in our country also considered. During the manufacturing process the development of initial crack generation and propagation of the crack to its critical value is considered in this work as a function of time of operation. In addition to that, Log-Normal distribution method is applied here for the assumption of the preexisting cracks in the manufacturing material while, considering the operational stresses and fatigue load the failure probability is calculated in the presented work. Variation with the initial crack size and various design stress is considered for this analysis.

This note presents a nonlinear control approach using dynamic neural network (DNN)-based feedback linearization (FBL) for nuclear reactor power control. The reactor model adopted in this study is based on neutronic dynamic and thermal-hydraulic models. The nonlinear plant is identified by a single-layer DNN trained using Quasi-Newton and Interior-Point methods. The feedback linearization scheme is combined with a Proportional-Integral (P-I) controller and simulations show good performance of the proposed controller. The efficacy of the controller is evaluated in the load-following mode of operation. Moreover, the fault-tolerance performance of the proposed approach is tested.

The self-regulating and self-stabilizing natures of a Pressurized Water Reactor (PWR) make it most suitable for the nuclear power industry. However, handling the disturbance and stabilizing a PWR with the disturbance are highly challenging tasks for control engineers due to inherent nonlinearity in a reactor. It leads to change in the behavior with variation in reactor power. Further, plenty of uncertainties exist in a PWR due to the heat transfer from fuel to coolant and the reactivity changes due to component faults and variation in fuel and coolant temperatures. Thus, it is essential to design a robust controller which can stabilize the system in presence of uncertainties and disturbances as well. In this paper, a state space model has been considered using the point kinetics equations of PWR coupled with the Mann’s thermal-hydraulic equations. The system matrices have been evaluated at different power levels with uncertainty in parameters to produce an interval state space model. A H ∞ based Full State Feedback Controller (SFSC) has been designed for this interval plant and then used for establishing a stability criterion in presence of disturbances. The outcomes have been validated using MATLAB simulations and discreetly exemplified in the result section.

Many pieces of research on the high efficiency and energy saving for power plants mainly focus on thermal power plants or pressurized water reactor nuclear power plants. There are few studies on the efficiency of the sodium-cooled fast reactor (SFR). To determine the energy loss during the operation for the thermal system of the supercritical reactor and design a better construction scheme for the supercritical reactor, in this paper the neutron reactor and the important equipment of the main heat exchange loop in CEFR were modeled, and the exergy loss and exergy efficiency during the operation under rated conditions are also analyzed. The results show that the biggest exergy loss of the main heat transfer system is the reactor core during CEFR operation. In the main heat exchange system of CEFR, the exergy efficiency of the once-through steam generator is the lowest.

This work presents a dynamic neural network-based (DNN) system identification approach for a pressurized water nuclear reactor. The presented empirical modelling approach describes the DNN structure using differential equations. Local optimization algorithms based on unconstrained Quasi-Newton and interior point approaches are used in the identification process. The efficacy of the proposed approach has been demonstrated by identifying a nuclear reactor core coupled with thermal-hydraulics. DNNs are employed to train the structure and validate it using the nuclear reactor data. The simulation results show that the neural network identified model is sufficiently able to capture the dynamics of the nuclear reactor and it is suitably able to approximate the complex nuclear reactor system.

This paper proposes a function that can interpolate real data as well as preserving the nature of the data. A rational cubic Bezier function expressed in terms of Ball control points and weights is presented. Four shape parameters are proposed which provides more freedom to the user to control the shape of the data. Sufficient conditions needed to preserve its positivity are presented. The degree of smoothness attained achieved $C$ 1 continuity. Three different real data are used in the application which are real data of oxygen level in gas, real data of a conductance obtained from experiments and real data consists of the output of a pressurizer inside a Pressurized Water Reactor. The output measures the water level of the pressurizer when the inputs which are the temperature and heat loss are varied. The results show that the proposed method managed to produce smooth interpolants as well as preserving the positivity of the data.

This work combines the subspace predictive control technique with the integral sliding mode control strategy to formulate a novel robust subspace predictive control scheme. The subspace predictive controller provides the nominal control whereas the integral sliding mode controller gives the discontinuous control action. The aim is to improve the capability of subspace predictive controller in handling uncertainties and external disturbances. The proposed control scheme is evaluated with a simulated pressurized water nuclear reactor. The effectiveness of the proposed technique is demonstrated for two different load-following operations in the presence of uncertainties.

Power grid fault will cause the frequency and voltage fluctuation of the nuclear power plant, which will directly affect the operation status of the coolant pump and the heat dissipation blade of the nuclear power unit. When the fluctuation degree exceeds the maximum allowable range of nuclear power equipment, may cause the nuclear power plant damage. In order to study the operation status of large-scale nuclear power plant equipment under grid failures, this article focuses on nuclear power plants in a certain area. A 500 kV grid model including pressurized water reactor(PWR) nuclear power plants is built on the PSCAD/EMTDC platform. The custom details of nuclear power plants are detailed The model includes reactor core, steam generator, steam turbine module, etc. The model can reflect the internal dynamic characteristics of nuclear power plant equipment in the event of a grid failure. By simulating the impact of voltage and frequency disturbances on nuclear power plants in the case of grid faults, the disturbance tolerance threshold for the safe operation of nuclear power plants is given, and equipment protection measures in the event of a three-phase short-circuit fault in the power grid are proposed, which is the fault ride-through of nuclear power plants. Further research on capabilities has laid the foundation.

According to the requirements of design and construction rules for mechanical components of pressurized water reactor nuclear islands, the primary hydrostatic test must be carried out before and after the operation of nuclear power plant. The overpressure protection system is used to prevent the accidental overpressure of the primary circuit during the hydrostatic test, and to alleviate the serious consequences of overpressure on the involved test equipment. The system design and equipment selection for the primary circuit of the third generation nuclear power plants are much different from that of the second generation ones. In addition, the construction of the third generation nuclear power plants is too late to have the reference experience. Based on the control characteristics of the primary process system in European pressurized water reactor and the deficiency of the overpressure protection system in china pressurized water reactor, the research work is conducted from the hereafter aspects, the pressure control strategy of the primary hydrostatic test, scheme design of the overpressure protection system, commissioning and site application. And it is also applied in the primary hydrostatic test of a nuclear power plant. The practical results show that the implementation principle of the overpressure protection system is correct, the analysis of risks and various countermeasures is appropriate, and the implementation scheme is feasible. The research work and practice can provide some reference for the subsequent project construction.

Spacer grids are essential for mechanical performance along the rods of the core. Spacer grids are always in a systematic square shape. It is significant to enhance the features of the grid with the vanes since the sublime coolant mixing and less pressure drop and must be permitted to the thermal hydraulic model. Researchers have used Computational Fluid Dynamics (CFD) in 3D Reynolds Averaged Navier-Stokes (RANS) and k-e turbulence model study with ascendable wall function skillfully as a supporting means in the evolution of spacer grids along the rods. Korean Atomic Energy Research Institute (KAERI) MATiS-H facility performed CFD evaluation of a (5x5) bundle of rods experiment which is submitted. Simulations were made in OECD/NEA CFD Standard application. The effect of turbulence in spacer grid geometries in a PWR (pressurized water reactor) is offered and also the impacts of results are evaluated. At reactor stream features and reactor close power three dissimilar vane arrangements are simulated and average thermo-fluid dynamic parameters are assessed. To simulate the characteristics of the flow among the rods and vanes, a tetragonal mesh is created. The simulations are performed with the CFD code of CFX 15.

A nuclear power plant is instrumented with a variety of sensors that monitor its variables to estimate the state and initiate safety actions, if needed. We address the problem of estimating drifts or errors in sensor measurements due to factors such as calibration changes. We propose an information fusion method that uses measurements from other sensors to generate an estimate of a sensor measurement, and its difference from an actual measurement provides an error estimate. We present two fusers based on the ensemble of trees and support vector machine that are trained using sensor measurements collected at an emulated test loop of a pressurized water reactor under no-drift conditions. We present error estimates for a differential pressure sensor of the heat exchanger of the primary coolant system, under twenty controlled scenarios using the test loop. Both positive and negative errors are captured by both methods in scenarios involving calibration drifts, blocking, air gap and electromagnetic interference. The root mean square errors of estimated drifts are typically within 2% percent of the maximum.

Good control of the pressure and water level of the pressurizer is a key prerequisite for the safe and economic operation of the pressurized water reactor (PWR) nuclear power systems. In this paper, the dynamic characteristics of pressure and water level of the pressurizer are analyzed based on marine PWR simulator firstly. It is found that its pressure and water level have obvious coupling characteristics. Then the least square method is used to identify the transfer function model of pressure and water level. A comparative analysis with the simulator data shows that the transfer model has a high degree of coincidence. Finally, a multivariable control method of pressure and water level based on MPC is proposed. The simulation results show that the method can achieve stable tracking of the target value accurately under the simultaneous change of pressure and water level setting.

The comprehensive analysis of advantages and limitations of topping nuclear power plants (NPP) with gas turbines (GT) is presented in the paper. The special accent was put on possibilities of enhancing the NPP steam cycles efficiency and problems connected with high cooling water demand. Moreover, the environmental impact of applying this technology in Polish power system was investigated, especially its potential to reduce the CO 2 emission. The basis for conducted investigation was formed by computer simulations carried out with Ebsilon Professional software. The performance of topped NPP steam cycle was also thoroughly compared with parameters that can be achieved in average Combined Cycle Gas Turbine (CCGT) power plant and NPP with pressurized water reactor (PWR). Eventually, the recommendations for further research and development of NPP steam cycles topped with gas turbines were made.

Compared to a conventional thermal power unit, a Pressurized Water Reactor (PWR) nuclear power unit is more prone to overspeed for its lower steam parameters. Disturbance from the grid may cause a significant rise of the rotor speed and even damage the nuclear reactor. Therefore, the installation of Overspeed Protection Control (OPC) is necessary, which functions by shutting the valve fast and decreasing the amount of the steam to limit the speed rise. Further researches on nuclear power units with OPC are important to the protection of nuclear power units. Practically, different OPC strategies can be obtained by switching the trigger signals or by controlling the valves selectively. Based on User Defined Model (UDM) in Power System Analysis Software Package (PSASP), a PWR nuclear power unit model with OPC is established. Different OPC strategies are analyzed when a fault occurs in the main grid. The simulation result verifies the protective effect of OPC, and it performs more effectively when the high/medium pressure valves and the trigger signals of power deviation and frequency deviation work complimentarily in a certain order.

According to feedback of Japan's Fukushima nuclear accident, beyond design basis earthquake would cause disastrous consequences for nuclear power plant. Referring to lessons learnt from Fukushima nuclear accident and nuclear power plant safety requirements, the seismic margin analysis (SMA) evaluation method was used for beyond design basis earthquake assessment for Pressurized Water Reactor nuclear power plants. According to the basic idea of FA method (fragility analysis method) and empirical data, SMA analysis of piping was explored in this paper by taking containment spray system piping as an example. The analysis result shows that the HCLPF (high confidence and low probability of failure) seismic level is 30.54 ft/s2, meet the requirements of review level earthquake 10.74 ft/s2.

Self-regulating and self-stabilizing nature of Pressurized Water Reactor (PWR) makes it fit to perform load-following operation. The dynamic model of a 100MWe PWR is presented which is assumed to be operating in Normal Mode. Optimal Nonlinear Dynamic Inversion (NDI) is adopted to design the load-following controller in such a way that the designed controller will guarantee global asymptotic stability along with perfect load-following operation and that too with least-possible control effort i.e. with minimum control rods movement. The controller performance has been validated using real-time Hardware-in-Loop (HiL) test-bench.

In the paper authors describe proposition of design and verification procedures of the discrete Fractional Order PID (FOPID) algorithm for control of the Pressurized Water Reactor (PWR) thermal power near its nominal operating point. The FOPID algorithm synthesis consists of: off-line optimal tunning of its parameters in continuous time-domain with LQ (Linear Quadratic) performance index and simplified models of nuclear reactor and control rods drive; its transformation into equivalent integer order structure with Oustaloup filters; and finally its transformation into equivalent discrete form. Discrete FOPID algorithm is further implemented in the PLC controller and verified by real-time simulation in the Hardware In the Loop (HIL) structure with non-linear nuclear reactor model. Promising simulation results were obtained, which confirm improved flexibility of the discrete FOPID algorithm in comparison to its classical PID counterpart.

The BOSS which is also called the set-on branch connection is a new structure used in nuclear power plant in recently years. The BOSS weld failure may lead to borated water leakage, and it will probably affect the safety of the equipment in the pressurized water reactor nuclear power plants (PWRs). In this study, causes of BOSS weld failures were studied first. Then through the technique features of radiographic inspection, an applicable inspection technique for BOSS welds in PWRs was summarized. Combining the experience feedback of BOSS weld failures at home and abroad, a useful maintenance solution was proposed. In the end, some feasible suggestions for improving the quality of BOSS welds were established from the point of views of design process, welding process and inspection process.

International audience; This paper proposes an innovative evolution of the current Pressurized Water Reactor (PWR) Nuclear Core Control strategies in order to satisfy more and more stringent maneuverability requirements. The following new control objectives are introduced: the hot and cold leg temperatures in contrast to the Average Coolant Temperature (ACT) in standard PWR control designs. This is made possible thanks to the use of the Primary Coolant Flowrate as a new control input. The proposed control design strategy first considers a point kinetic nuclear core non-linear model, calibrated on a highly realistic nuclear core simulation code. Secondly, it proceeds to a multi-objective H 2 /H ∞ control synthesis, using an adequate control structure and non-smooth optimization.

The detailed models, which indicate the dynamic processes of mega-kilowatt pressurized water reactor (PWR) nuclear power plant (NPP), have very high order. They are difficult to be appied to the medium-to-Iong term simulation of power system, so the reduction of their order is of great value. A new strategy, which deals with the physical thermal circuit and the control system separately, is proposed in this paper according to the structure and operation of PWR. Then, the dynamic responses, with and without the influence of main pump, are carefully studied using the detailed models. It is shown that only the change of the flux is needed to consider particularly in the modeling of coolant circuit. Finally, the method of reduction and identification is presented for each individual link of the original detailed models.

In a typical pressurized water reactor's (PWR) steam generator, the primary component for distribution of feed water in secondary side is feed water ring. In this research, computational fluid dynamics analyses are carried out to determine the flow as well as pressure distribution in feed water ring with J tube. Furthermore, velocity contour at the J tube and feed water ring junction has been plotted. It was concluded that the junction is subjected to erosion-corrosion because of a higher erosion velocity through those tubes. So, by moving these tubes to other locations can increase the life of feed water ring. The flow distribution of feed water ring of Chashma nuclear power plant was also compared with other kinds of headers used. A modified design of feed water ring was suggested. The modifications resulted in a decreased erosion-corrosion risk on the surface of feed water ring. Additionally, the performance of feed water ring in main feed water line break accident was also carried out, reverse flow observed before closure of check valve was 1154 kg/sec. A pressure surge of 52 MPa was observed after the closure of check valve, which may result in water hammering and can damage the feed water ring.

The Rancor Microworld is a recently developed low- fidelity Nuclear Power Plant Simulator. The microworld concept provides a simplified representation of an integrated 30 MW pressurized water reactor that can be taken from a cold-shutdown through turbine startup and load generation. Non-expert operators can be trained to competently operate the Rancor simulator within an hour, and it has already been used to examine attention and situation awareness measures in a series of experiments. Here we discuss the role of simulation in nuclear human factors, examine how a tool like Rancor fills a niche for nuclear human factors process control, and discuss potential avenues for future development and research in computer-based procedures, human reliability analysis, team operations, and cybersecurity.

In this paper a strategy for an adaptive gain scheduled PID controller is proposed for PWR type of Nuclear reactor. Nuclear reactors are highly nonlinear in nature. The proportion of Pressurized Water Reactor is beyond 60% of the total Nuclear reactor components, since PWRs are the most desirable reactors for producing nuclear power. The reactor power level is regulated by adjusting the control rod speed. The reactor model is developed from the basic point-kinetics equations with reactivity feedback and thermal hydraulic considerations. The reactor model is linearized at different percentages of reactor power levels. There is a significant variation in linear model parameters and Gain scheduled adaptive PID control scheme is attempted to accommodate the variations in model parameters at different power levels. The performance of proposed controller is compared with fixed PID controller. From the results obtained it is found that the proposed control scheme gives an improved performance over conventional PID.

VVER-1200 NPP (Nuclear Power Plant) is a 3+ generation Russian designed PWR (Pressurized Water Reactor). This type of NPPs will be built to increase the capacity of Paks NPP. In our paper the impact of the mixing vane of VVER-1200 spacers on the flow is examined. Distribution of velocity, temperature field and cross-flows are evaluated. For these investigations detailed calculation models have been developed with the ANSYS CFX code. A mesh sensitivity analysis is performed using three meshes of different resolutions. The impacts of the spacer orientations and arrangements on the mixing between two fuel assemblies are analyzed.

Constructon of optical lighting systems with LEDs means the use of non-imaging optical elements with different shapes refractive and reflective surfaces. Well-known calculation methods, which are used for the calculation of spherical and similar to it aspheric optics, are not applicable. By itself, the emitting area of ??the LED emits according to the law, which is close to Lambertian. Typically, LEDs are already equipped with the primary lens. They are used in lighting units to illuminate the fuel elements which is extremely advantageous when the light output is used as fully as possible. In the study, the technique of a fundamentally new design of the lighting unit. The main element of lighting system will become LED.

Majority of fission power reactors in the world are pressurized water reactor (PWR) type. Burning UO2 fuel in PWR produces long lived radioactive isotopes like minor actinides. The characteristics of minor actinide transmutation in specific PWR fuel assembly are studied. The feasibility of using minor actinides as burnable absorber for transmutation and reactivity compensation is studied. The fuel assemblies with conventional burnable absorber rods are replaced with minor actinides rods. Selected assemblies are burnt for 720 full power days, having same initial reactivity for both burnable poison and minor actinide cases. The calculated results indicated that the assemblies with minor actinides yield less cycle length as compared to assembly with conventional burnable poison rods. However a significant transmutation of minor actinides reduces the radiotoxicity of spent fuel to considerable magnitude.

In the paper the multi-nodal Pressurized Water Reactor (PWR) heat transfer model based on Mann's model is presented. This model is used for heat transfer from fuel to coolant modelling purposes in a reactor core. The authors expand the approach widely used in literature by defining additional coefficients for heat transfer model. These parameters approximate the power generation distribution in the PWR reactor core according to the reactor control rod bank movement. Authors describe in details the proposed methodology for calculation of introduced power distribution coefficients. In the simplest case, those coefficients may be equal and constant over time and space of a reactor. On the other hand, they can be treated as variables that are time and reactor region dependent. By introducing these specific coefficients, the nodal model of heat transfer gains advanced capabilities that can be efficiently used in design and synthesis of more advanced and complex power control system in nuclear reactor with respect to temperature distribution within the reactor core.

Nuclear plant safety as well as availability can be quite well improved if such functions as early fault diagnoses and on-condition maintenance are automated. To realize these functions, on-line monitoring of nuclear reactors is required, and state observation is one of the key and basic techniques of reactor-state monitoring. Moreover, state-observation is also the basis of applying advanced reactor control strategies. State observation deals with recovering desired states from available measurements. Pressurized water reactor (PWRs) are complex dynamic systems with high nonlinearity and strong uncertainty, which leads to the necessity of developing nonlinear adaptive reactor observation technique. Motivated by this, a nonlinear adaptive observer with only the measurements of both the nuclear power and reactor inlet coolant temperature is proposed for PWRs in this paper. It is proved theoretically that this new observer is globally convergent and robust to the modeling and parameter uncertainties and exterior disturbances by a well-designed adaptation law. Numerical simulation results show the feasibility and high performance of this new nonlinear adaptive observer.

The thermodynamic efficiency of typical conventional and nuclear power plants is insufficient. A large amount of the generated heat is removed to the environment and regarded as a waste. A solution which is commonly used to increase the efficiency and profitability of the existing and planned power plants refers to cogeneration of heat and power. To prepare sophisticated thermodynamic and economic analyses, a precise static model of the steam/water cycle of the considered power plant is essential. The paper presents a methodology to build a static model of the steam/water cycle of the Nuclear Power Plant with Pressurized Water Reactor. The discussed model was prepared using phenomenological and network approach, complemented by a precise reference model as data source. The prepared model is believed to be very useful in thermal and economic efficiency analysis.

Pressurized water reactor (PWR) is the most widely used fission reactor whose safe, stable and efficient operation is needed by the renaissance of fission energy industry. Power-level control technique that strengthens both the closed-loop stability and transient performance is quite beneficial to improve the operational performance of PWRs. In this paper, based upon the physically-based approach, it is showed that general PD output feedback laws can guarantee the globally asymptotically closed-loop stability under a certain sufficient condition. Then, the influence of the gains in coolant temperature feedback loop to the regulation performance is shown by numerical simulation results, which is in consistence with the theoretic result. This work gives a thorough explanation to why classical PD output feedback control is effective for PWR power-level regulation practically, and also shows the feasibility of applying simple PD controller for deep load following.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.

GE-Reuter-Stokes (GERS) has developed a new line of neutron proportional counters, the B10Plus+ proportional counter. The detector design is intended to serve as a cost-effective alternative to traditional 3He proportional counters in a variety of applications. The detector is a hybrid design 10B-lined tube optimized with the addition of a small quantity of 3He gas to improve the detector performance and efficiency. As a demonstration of the B10Plus+∗ detector, GERS has constructed a Uranium Neutron Collar (UNCL) system consisting of B-10Plus+∗ proportional counters. GERS has designed and built a demonstration UNCL system intended to match the performance of a Type-I UNCL design in Pressurized Water Reactor (PWR) geometry operating in thermal mode. GERS offered their system on loan to the International Atomic Energy Agency (IAEA) Safeguards Division of Technical and Scientific Services for an assessment of the detector technology and the demonstration system. We have characterized the demonstration UNCL system and compared its performance with a traditional Type-I UNCL design in regular use by the IAEA. This paper summarizes our findings and observations during the characterization and testing activity.

A pressurized water reactor (PWR) nuclear power plant model (NPP) representing one unit of the Palo Verde Nuclear Power Generating Station is incorporated into Positive Sequence Load Flow (PSLF) software. The nuclear steam supply system (NSSS) model comprises a point kinetics description of neutronics and thermodynamics in the reactor core, pressurizer, plenums, hot and cold legs. In addition, the PWR considered here utilizes a recirculation (U-tube) steam generator. A sliding-average-temperature control program for the reactor, pressurizer pressure controller, and simple turbine and governor models are also developed. Subsequently, the NSSS representation is implemented into PSLF by user-defined modeling. Simulation results of isolated plant for a step change in reactivity due to control rod withdrawal are presented and then effect of a disturbance in the medium-sized power system on the developed model of NPP is evaluated.

Thorium can be utilized in Pressurized Water Reactors (PWR) as ThO 2 fuel matrix supported by fissile materials. In this analysis, VVER-1000 reactor was taken as a representative of the PWR reactor group. The thorium fuel matrix was mixed with fissile material in form of pure 233U, enriched uranium (19.7 % enrichment) and Reactor-Grade plutonium. Whole fuel cycles were analysed by ANDREA nodal code with diffusion data prepared by HELIOS calculations. Equilibrium fuel loadings were achieved for all fuel composition for four different lengths of fuel loadings reaching from 4×12 months to 5×18 months. The process of reaching the equilibrium loadings, fissile material requirements, and fundamental operational parameters are presented.

Pressurized water reactor (PWR) is now the most widely utilized nuclear reactor, and its safe, stable and efficient operation is meaningful to the current rebirth of nuclear fission energy industry. Power-level control is a crucial technique which can deeply affect the operation stability and efficiency of PWRs. Comparing with the classical power-level control, the advanced power-level control may strengthen both the closed-loop stability and control performance by feeding back internal state-variables. However, not all of the internal state variables of a PWR can be measured directly. To implement advanced power-level control laws, it is necessary to develop a state-observer to reconstruct the unmeasurable state-variables. Since a PWR is naturally a complex nonlinear system with its parameters varying with the powerlevel, fuel burnup, Xenon isotope production, control rod worth and etc, it is meaningful to design a nonlinear observer for the PWR with adaptability to system uncertainties. Due to this and the strong learning capability of the multilayer artificial neural network (MNN), an MNN-based nonlinear adaptive observer is given for PWRs. Based on Lyapunov stability theory, it is proved theoretically that this newly-built observer can provide bounded and convergent state-observation. This observer is then applied to the state-observation of a special PWR, i.e. the nuclear heating reactor (NHR), and numerical simulation results not only verify its feasibility but also show the relationship between the performance and parameter values of this newly-built observer.

The goal of the EDF TREFLE® project is to provide to the EPRTM project a full-scope simulator in order to perform design studies and operator training for the EPRTM reactor under construction at Flamanville, France. This article describes the 3D real time neutronic module, COX-SIMU, developed by EDF engineers and integrated in the simulator by the CORYS T.E.S.S Company. For the calculation of the 3D power distribution, COX-SIMU solves the neutron diffusion equations with 2 energy groups (transient and steady states). A full 3D distribution is needed due to the specifications of the EPRTM reactor internal instrumentation which requires knowledge of the detailed flux distribution for the Protection System. COX-SIMU includes thermal-hydraulical and thermal conduction modules to take into account the water moderator and fuel feedback on neutronic reactions. A simple model based on exponential functions is used for the calculation of the Residual Heat. COX-SIMU is fully validated against the industrial EDF neutronic code COCCINELLE with fine mesh discretization. The modularity and flexibility of COX-SIMU allow its integration in any Pressurized Water Reactor simulators. Thus, COX-SIMU is also integrated in the EPRTM TAISHAN Simulator.

A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The effect of the load-following operational mode on the key variables of the NSSS is observed.

Advanced small modular reactors (aSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. aSMRs are conceived for applications in remote locations and for diverse missions that include providing process or district heating, water desalination, and hydrogen production. Several challenges exist with respect to cost-effective operations and maintenance (O&M) of aSMRs, including the impacts of aggressive operating environments and modularity, and limiting these costs and staffing needs will be essential to ensuring the economic feasibility of aSMR deployment. In this regard, prognostic health management (PHM) systems have the potential to play a vital role in supporting the deployment of aSMR systems. This paper identifies requirements and technical gaps associated with implementation of PHM systems for passive aSMR components.

THE Jules Horowitz Reactor (JHR) [1] is the next Material Testing Reactor under construction in France. It will be a major research infrastructure in Europe to support existing power plant operation and lifetime extension, as well as future reactor developments. Its objectives are to participate to the demonstration of GEN-II vessel behavior for Pressurized Water Reactor lifetime extension, and to test the new structural material and fuel behavior under irradiation for the development of GEN-III and GEN-IV reactors. It also aims at producing a large part of the radioisotope European request.

IAEA verification of the nuclear material contained in fresh nuclear fuel assemblies is usually based on neutron coincidence counting (NCC). In the case of uranium fuel, active NCC provides the total content of uranium-235 per unit of length which, combined with active length verification, fully supports the verification. In the case of plutonium fuel, passive NCC provides the plutonium-240 equivalent content which needs to be associated with a measurement of the isotopic composition and active length measurement to complete the verification. Plutonium isotopic composition is verified by high resolution gamma spectrometry (HRGS) applied on fresh fuel assemblies assuming all fuel rods are fabricated from the same plutonium batch. For particular verifications when such an assumption cannot be reasonably made, there is a need to optimize the HRGS measurement so that contributions of internal rods to the recorded spectrum are maximized, thus providing equally strong verification of the internal fuel rods. This paper reports on simulation work carried out to design special collimators aimed at reducing the relative contribution of external fuel rods while enhancing the signal recorded from internal rods. Both cases of square lattices (e.g. 17×17 pressurized water reactor (PWR) fuel) and hexagonal compact lattices (e.g. BN800 fast neutron reactor (FNR) fuel) have been addressed. In the case of PWR lattices, the relatively large optical path to internal pins compensates for low plutonium concentrations and the large size of the fuel assemblies. A special collimator based on multiple, asymmetrical, vertical slots allows recording a spectrum from internal rods only when needed. In the FNR case, the triangular lattice is much more compact and the optical path to internal rods is very narrow. However, higher plutonium concentration and use of high energy ranges allow the verification of internal rods to be significantly strengthened. Encouraging results from the simulation pave the way to experimental confirmation of feasibility of the multi slot collimator concepts.

The coordination between nuclear power units and power grids is very important for the safe and stable operation of both nuclear units and power grids. Based on a complete mathematical model of pressurized water reactor (PWR) nuclear power plant (NPP) developed for power system electromechanical transient, mid-term and long-term dynamic simulations, the dynamic performance of the NPP with respect to grid voltage and frequency disturbance is analyzed, and the interaction and coordination between NPPs and power grids are studied. Thus, the principle and scheme of the coordinated protection and control between large-capacity NPPs and power grids are proposed. The results of this paper can be used as a reference for the optimization of parameters, the study of coordination between NPPs and grids, and their safe and stable operation.

Spacer grids are an integral part of pressurized water reactor (PWR) fuel assemblies. They provide structural integrity as well as play a vital role in defining the thermal hydraulic behavior of the fuel assembly. The spacer grid design is continuously being improved to achieve better heat transfer characteristics. The studies performed to date include the design optimization and effects of mixing vanes on the thermal hydraulic behavior of fuel assembly. Majority of these studies do not include the effects of springs and dimples of the spacer grid. The present study was performed to analyze the flow behavior and pressure drop across a single coolant channel in the region of spacer grid along with all geometrical details. Detailed analysis of a single coolant channel comprising spacer-grid cell with dimples and spring was done using Ansys-CFX. The mesh independency study was carried out to make a compromise between solution time and accuracy. The problem was solved using K-e model along with scalable-wall-function option. The Reynolds averaged mass conservation and momentum conservation equations were solved and the pressure and velocity field of flow channel was obtained. The average velocity varies from 2.52m/s to 7.35m/s for Reynolds number variation from 4.0×104 to 11.7×104. The cross flow velocities vary from 0.2 to 0.57m/s for these flow conditions. The Euler numbers calculated for these flow conditions were compared to the documented experimental results and were found to be in very good agreement. The analysis results conclude that single cell approach is an accurate and economical approach for predicting the pressure drop across the spacer-grid. The analysis also confirmed the applicability of k-e model with scalable-wall-function for the prediction of the single phase hydraulic characteristics of spacer-grids.

Aging plants and reduced workforces are causing commercial nuclear power reactors to limit the extent of their predictive maintenance programs to only critical to safety and critical to operation systems. Some sites are even forced into living with a reactive maintenance strategy for all but the most critical systems. This typically leads to significant expense through costly downtime and component unavailability. Advancements in wireless technology have resulted in the implementation of measurement devices on some important systems that have alleviated the condition monitoring and testing burden on maintenance personnel. Through research and development (R&D) sponsored by the Department of Energy (DOE) Small Business Innovation Research (SBIR) program, a wireless condition monitoring application has been identified inside the containment of a pressurized water reactor (PWR) that can benefit from the introduction of wireless technology. The application that has been selected is the use of wireless vibration sensors to monitor the condition of containment cooling fans. The host utility for this project is the Arkansas Nuclear One (ANO) power generating station. Due to the location of the fans within the containment building, maintenance personnel are prevented from manually collecting vibration data frequently enough (e.g., monthly) to determine the health of the equipment. Through the use of wireless sensors, data can be collected once a day to provide a more complete picture of the fans' health with advanced warning of a failing component. This paper will document the efforts of this project to deploy a wireless vibration system at ANO.

A pressurizer (PRZ) mechanism model was employed for pressurized water reactor (PWR) nuclear power plant to meet rapid development of its simulation requirements. According to PRZ operational characteristics, the laws of energy conservation, mass and momentum conservation are used in obtaining a dynamic two-phase imbalance mechanism model which considering the influence of the spray flow, the power of electric heater as well as factors related to the safety valves. Particle Swarm Optimization (PSO) Algorithm was proposed to obtain a set of optimized parameters in the PRZ mechanism model. The parameter optimization algorithm was applied to a simulation case in PRZ of a 900MW PWR nuclear power plant, and a pressure simulation model was created by means of module packaging. By comparing the simulation results with the datum obtained by the PWR nuclear power plant, the correctness of modeling and the effectiveness of optimization methods were verified.

When the working condition of primary pump of pressurized water reactor changes, the countercurrent of fluid can be well prevented by check valve. The water hammer phenomenon often appears in the loop during the countercurrent process. Serious water hammer not only brings incident of over pressure, and imperil pressure boundary, but also engenders the lapse of check valve. Contra-push check valve is a new kind of check valve. Its new structural design of river diversion can well reduce the water hammer phenomena, thereby enhancing the operational security of nuclear power plant. By the aid of the numerical methods, physical models, dynamic mesh model and UDF(User Define Functions) offered by FLUENT, simulation of the local irregular flow field and the pressure distribution during the process of closing the valve have been conducted more realistically. Simulation results indicate that, the change of geometry shape will cause the higher local velocity than the other regions but little affect of stability during shutdown. The analysis of results are providing a basis for future applications and reliability analysis of valve.

The recent Fukushima accident shows that for nuclear power plants, station blackout (SBO) is an important cause of core damage severe accident and radiological releases. In this paper, the typical three-loop pressurized water reactor (PWR) in China is modeled using an integrated severe accident analysis code, and the progression of the SBO accident sequence is also simulated. Then based on the analysis of the calculation results, several measures are adopted to mitigate the SBO accident. In spite of some negative effects, the analysis results show that the consequences of SBO severe accident can be effectively mitigated by restoration of the AC power supply, rapid depressurization of the primary system by opening the three safety valves, and feeding water into the steam generators.

Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20–30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory — containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our belowground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino detectors that were deployed. Finally, some preliminary results of our aboveground test will be shown.

Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time.