Operation Of Nuclear Reactors Research Articles

Research for axial power distribution control of a space nuclear reactor based on nonlinear model

During the operating process of a space nuclear reactor, once the local overload of heat energy occurs in the core, the corresponding fuel elements would be damaged and the lifespan of the reactor would be seriously reduced. In fact, the heat flux at any point in the core is proportional to the neutron flux density. A sufficiently flat axial neutron flux density distribution can protect the fuel assembly from frequent load changes. It is so clear that the axial power distribution control is crucial for the safe and stable operation of the space nuclear reactor. However, the space nuclear reactor is a distributed parameter system, which brings great challenges to the traditional lumped parameter control methods. In this paper, a distributed parameter control method is proposed to suppress the initial neutron flux peaks and realize the exponential stability of the neutron flux distribution control system. First of all, the spatiotemporal model of the reactor core based on the neutron diffusion equation is established. Then, the stability analysis of the control law for the neutron flux distributed parameter control problem is performed by constructing the Lyapunov function. The optimization technique based on the spatial differential linear matrix inequality is used to approximate the solution of the suboptimal integral control law. The control law is implemented by rotating drum, and the flat distribution of the axial power of the core is realized by the distributed parameter control. Finally, the numerical results are provided to validate the feasibility of the proposed control scheme.

Identification of the need for unmanned radiation monitoring facilities in the radiation monitoring of research reactors in Indonesia

The facility’s licensee conducts environmental radiation monitoring in nuclear facilities to monitor radiation exposure in the facility’s vicinity. This activity is carried out also to monitor radiation release as a result of nuclear reactor operation. Aside from that, monitoring also works as a device to monitor radioactive release in a nuclear emergency. Therefore, the radiation monitoring system is crucial in nuclear utilization facilities to determine the number of radiation exposure to the surrounding environment. However, the existing stationary monitoring system has a risk of being unable to work if the system is down in case of natural disaster occurs. One way to mitigate this risk is to deploy an unmanned radiation monitoring system to monitor radiation exposure without putting personnel at risk. To define a suitable unmanned radiation monitoring system, identification of facilities and infrastructure required to design an unmanned radiation monitoring system for a research reactor in Indonesia is carried out. Facilities and infrastructure needed for unmanned radiation monitoring systems are unmanned aerial vehicles, radiation detector, control and communication module, navigation system, and software for the control system. These required facilities and infrastructure are then specified to determine the necessary specification for monitoring research reactor in Indonesia. The facilities’ required specifications are unmanned aerial vehicles with rotary-wing type, CdZnTe Detector, and GPS/GLONASS based navigation system. For infrastructure specification, control and communication module and software for the control system is not specified in how the system could meet the expected required performance rather than in detail. However, the system must provide and process measurement data in real-time to be presented in a radiation heatmap. Keywords: Identification, Radiation Monitoring, Unmanned

A Versatile Methodology for Reactor Pressure Vessel Aging Assessments

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime threshold. From the perspective of possible further lifetime extension, satisfying safety requirements is a priority. Characterization of the structural integrity of the reactor pressure vessel (RPV) is an important issue as it is a guiding parameter that influences the reactor lifetime. Embrittlement of RPV material is primarily induced by the bombardment of fast neutrons (with energies greater than 1 MeV). Consequently, fast neutron fluence is one of the quantities used by safety authorities to characterize the structural integrity of RPV. However, future RPV aging assessments might lean on new variables with respect to current laws, such as neutron fluence considering the whole neutron spectrum or displacements per atom (dpa) since the latter is more representative of overall damage generated in the RPV. In order to meet these challenges, a versatile calculation scheme for RPV aging assessments is proposed in this paper. The developed methodology allows one to compute (fast and non-fast) neutron fluence as well as dpa rate, using the Norgett-Robinson-Torrens dpa model and the Athermal Recombination Corrected dpa model, for a wide azimuthal and axial range on the RPV and in the capsules of the aging monitoring program (which contain dosimeters and vessel material samples). This methodology is based on a coupling between deterministic (CASMO5 and SIMULATE5) and Monte Carlo (MCNP6) numerical approaches. First, the deterministic approach is used to evaluate the full-core fission neutron source term. Second, Monte Carlo modeling is used to perform the neutron attenuation from the core to sites of interest, such as the RPV. The computational efficiency, accuracy, and potential benefits of the methodology are presented. Moreover, the frequency at which neutron transport calculations should be performed in order to obtain sufficiently accurate time-integrated data over a reactor cycle is discussed. Finally, the validity of the fast neutron fluence as an indicator of RPV aging is compared against the use of dpa.

Prediction of safety parameters of pressurized water reactor based on feature fusion neural network

Rapid and accurate calculation of safety parameters based on the state and layout of the core is essential to ensure the safe operation of nuclear reactors. Based on artificial intelligence technology, we established a neural network model with seven layers, called the feature fusion neural network (FFNN). A residual compensation layer is added in the third hidden layer to avoid gradient dispersion, and a feature fusion layer is added in the last hidden layer to fuse low-dimensional features and high-dimensional features. Numerical experiments simulate the pressurized water reactor of Qinshan Nuclear Power Station, and 10,000 sets of safety parameters are calculated by CASMO5 under different arrangements as ground truth. Experiments show that the mean relative error of the subassembly power peaking factor and rod power peaking factor estimated by FFNN is about 1%, and the maximum relative error of the limited effective multiplication factor is within 0.5%.

Operation and safety analysis of space lithium-cooled fast nuclear reactor

As the power requirements for deep space exploration continue to increase, space nuclear power systems are imperative. In this paper, a set of models are established for the lithium-cooled space reactor combined with the Stirling engine. The space lithium-cooled reactor system is modeled, and the result of the steady-state is checked with the maximum relative error of 13.3%. Moreover, the characteristics under the unprotected reactivity insertion accident (URIA) and the unprotected loss of heat sink accident (LOHA) are obtained and analyzed. The results showed that: a) the solar heat flux cause the radiator temperature to fluctuate but has limited impact, b) under LOHA, the temperature of hot spot decreases rapidly from 1436 K to 1423 K, verifying the inherent safety of the system, c) under URIA, the hot spot temperature rises to 1546 K within 160 s. This work provides a solid basis for the design and analysis of space nuclear systems.

Topology based multiphase flow model for nuclear safety applications

Computing the working conditions and transient processes of nuclear reactors is important for both the safety and design of these reactors. The operation of nuclear reactors depends on various phenomena resulting from different flow regimes in the core. To accurately describe the system’s operation, multiphase, multi-scale models should be integrated for dispersed and separate flow topologies. This work presents a new multiphase flow model which accounts for the actual flow topology (dispersed or separate) and provides a unified model for all flow regimens. The basis of the new model is a physical insight that guides the averaging process of the multiphase conservation equations. The study defines a topology function for each fluid which expresses its connectivity on the scale of the computation grid. The new topological function is used to divide the mixture’s averaged conservation equations into phasic equations in a way which correctly accounts for stress and heat transfers in the mixture. From this perspective, we show a unified model which is able to correctly describe the interactions between the phases in the flow field. This model can provide the flow starting from dispersed conditions and moving to separate flows.

Marangoni Convection of Dust Particles in the Boundary Layer of Maxwell Nanofluids with Varying Surface Tension and Viscosity

The flow of nanofluids is very important in industrial refrigeration systems. The operation of nuclear reactors and the cooling of the entire installation to improve safety and economics are entirely dependent on the application of nanofluids in water. Therefore, a model of Maxwell’s dusty nanofluid with temperature-dependent viscosity, surface suction and variable surface tension under the action of solar radiation is established. The basic equations of momentum and temperature of the dust and liquid phases are solved numerically using the MATLAB bvp4c scheme. In the current evaluation, taking into account variable surface tension and varying viscosity, the effect of dust particles is studied by immersing dust particles in a nanofluid. Qualitative and quantitative discussions are provided to focus on the effect of physical parameters on mass and heat transfer. The propagation results show that this mixing effect can significantly increase the thermal conductivity of nanofluids. With small changes in the surface tension parameters, a stronger drop in the temperature distribution is observed. The suction can significantly reduce the temperature distribution of the liquid and dust phases. The stretchability of the sheet is more conducive to temperature rise. The tables are used to explain how physical parameters affect the Nusselt number and mass transfer. The increased interaction of the liquid with nanoparticles or dust particles is intended to improve the Nusselt number. This model contains features that have not been previously studied, which stimulates demand for this model among all walks of life now and in the future.

A A Pilot Review of State-of-the-Art Deep Learning Applications to Nuclear Engineering & Technology

Given the advent of big data (BD), the focus of this “pilot” review of the most recent literature is upon deep learning (DL) implementations/solutions to important, as well as inherently complex, nuclear engineering and technology (NET) problems. The potential advancement from implementing large/real-scale DL-based algorithms is critically detailed, as well as highlighted, by inspecting the relevant body of latest research which appears in two major – overlapping, yet also complementary to each other in terms of scope and audience - nuclear energy (NE)-dedicated scientific research means of publication, namely Elsevier’s “Annals of Nuclear Energy” and “Progress in Nuclear Energy” journals. It is demonstrated that, given their (I) accuracy, robustness-to-noise and resilience to incomplete/erroneous information, as well as (II) competence under non-stationary and transient nuclear reactor (NR) operating conditions, deep and convolutional neural networks (DNNs and CNNs, respectively), as well as their variants, are becoming increasingly important, in their own right, for the swift and reliable solution of NET problems. It will be of great interest for – while also advantageous to - the NE, BD, and computational intelligence (CI)/DL communities, to join forces, follow up, and contribute to future research in – as well as application to - this particularly interesting and fruitful collaboration between disciplines.

Advanced frequency-domain compensator design for subsystems within a nuclear generating station

Like all electrical generating stations, nuclear plants possess many feedback structures. Application of frequency-domain control design techniques has the potential to improve the performance of these feedback systems, thus improving the overall resiliency of the generating station and the grid at large. Typically the applicability of such design techniques are limited by the availability of detailed plant descriptions. In this work, an analysis is performed of several control structures of a nuclear generating station from a frequency-domain control perspective. The model used is that of a training simulator for reactor operators. The applicability of high performance control techniques to a number of subsystems is assessed and compensators are developed for the most promising systems. In recent years there has been renewed work in the application of classical control techniques to systems where the plant model is not fully known. One such technique is the Cloud technique, which is used to develop a type of controller known as a Modified Bode Optimal (MBO) controller. This technique is applied to develop large feedback compensators for two control systems within a simulated nuclear generating station. The relative performance of these systems is assessed using traditional time-domain metrics.

Application of model free active disturbance rejection controller in nuclear reactor power control

The power control of nuclear reactors is the key to the safe operation of nuclear reactors. At present, PID controllers are still used in most industrial control. Although PID has the advantages of simple design and easy parameter tuning, it is still difficult to overcome its own shortcomings in the control of complex systems and large-scale parameter adjustment. Model-Free active disturbance rejection control(MF-ADRC) and PID are control method that do not require model information. This study compares the control performance of PID and MF-ADRC in nuclear power control. First, three different core models are established, and the power MF-ADRC and PID controller of LBFR is designed. The performance of MF-ADRC and PID is compared through reactive disturbance and wide-range power adjustment. Finally, the performance of MF-ADRC and PID control is compared through the control of different core power. The results show that MF-ADRC has the same parameters to be tuned as PID; PID is difficult to balance overshoot and settling time, but MF-ADRC does a good job in this aspect; PID and MF-ADRC have good robustness, In terms of results, MF-ADRC is more satisfactory.

A modeling error compensation approach for the feedback control of the nuclear reactor operation

Nuclear energy for civil applications has been a matter of renewed interest as it offers clean, safe, and reliable energy sources. A new generation of compact reactors and improved safety strategies have driven refreshed research on the design and control of nuclear plants. New results on the regulation of the power output of nuclear reactors should provide improved operation strategies oriented to reliability and safety. The present work aims to study the stability and performance of a nuclear reactor under nonlinear feedback control actions. The study considered the five states March-Leuba mathematical model of a boiling water reactor. A nonlinear feedback controller that leads to exact cancellation of nonlinearities was obtained by dynamical inversion methods. However, uncertain parameters and unmeasured signals are commonly found in practical conditions. To address this issue, an observed-based estimation scheme was proposed to obtain estimates of the unmeasured signals and the trajectory of the modeling error, which are subsequently used to obtain an approximate version of the exact feedback inversion. Numerical simulations for open-loop stable and unstable (e.g., large-amplitude limit cycle) showed that the proposed control scheme provides closed-loop stability despite large model uncertainties.

Error compensation and flexible measurement of nuclear fuel assembly detection device for film thickness and morphology deformation

Regular detecting film thickness and morphology deformation of nuclear fuel assembly (NFA) is an essential measure to ensure safe operation of nuclear reactors. This paper presents an error compensation and flexible measurement method for an NFA detection device to realize safe, accurate and efficient measurement of film thickness and morphology deformation. The detection device combining adaptive aligning mechanism, flexible measuring mechanism, and serial-parallel hybrid measuring mechanism is designed for accurate contact measurement of complex deformed NFA. By probabilistically adjusting compensation parameters of the closed-loop measurement loop constructed by serial-parallel hybrid measuring mechanism, a comprehensive error compensation method is proposed to further improve measurement accuracy. Through dynamically monitoring the contact measuring forces calculated by mechanical model without additional sensors, a flexible measurement method is proposed to ensure safe detection of high-risk NFA. Experimental results indicate that the device after applying the proposed methods can complete accurate and flexible measurement of NFA. The maximum measurement errors of film thickness and morphology deformation are reduced to 3.5 μm and 8.3 μm respectively, and the maximum measuring forces are controlled within 4.37 N.

Analysis and evaluation of the irradiated dimensional and volume change behavior of nuclear graphites

Literature on the changes in the specimen dimensions and volume after high energy neutron irradiation of more than a dozen graphites were examined. Some of these data have been previously used to design graphite components in gas cooled reactors. The dimensional change behavior of a graphite is rather unique and depends on both the irradiation temperature and accumulated irradiation dose. The characteristics of the dimensional and volume change behavior of several graphites were analyzed and evaluated from engineering design consideration of graphite components. Specifically, six parameters were derived from the dimensional and volume change behavior. These parameters can serve as “figures-of-merit” (FOM) to enable engineering judgement on the irradiation-induced dimensional and volume change behavior of nuclear graphites, and are referred to as “derived figure-of-merit (DFOM).” The DFOMs also provide engineering bases to examine the opportunity to select several competing graphites for fulfilling the design and operation of components in reactor service. The utility of these DFOMs can be realized potentially in designing inservice inspection and component condition monitoring (CCM) protocols that are effective and efficient both technically and economically. The DFOMs, in combination, also can guide the reactor operator to arrive at optimum component replacement schedule for graphite components, which ensure maintaining graphite components functional requirements and related reactor safety issues.

Generation of hydrogen by hydroheterogeneous compositions based on aluminum and alkali metals

The process of hydrogen formation and the associated risk of combustion and explosion is a complex problem concerned with the hydrogen and radiation safety of nuclear reactors. Lithium, potassium and sodium hydroxides are used in VVER reactors as corrective additives for keeping the hydrogen potential of the water coolant with boric acid at a controlled level of 5.8 to 10.3. In the process of investigating the interaction of aqueous solutions of the above hydroxides with aluminum, the most chemically active of these is lithium hydroxide; this reaction proceeds with hydrogen formed at a high rate at room temperature (in an exothermic mode). The processes of hydrogen generation in hydroheterogeneous compositions with potassium and sodium hydroxides proceed at an acceptable rate with heating to ~ 60 °C. The kinetics of hydrogen generation depends in a complex way on the content of boric acid, namely, the hydrogen yield is at a level of ~ 1000 ml at a low concentration of 0.01 to 0.05 g/l, and there is no hydrogen formation at a concentration of 0.6 g/l. According to the coolant quality standards, in the hot state of a VVER-1000 unit or in the reactor state at the minimum controlled power level, the total concentration of alkali metals is about 1 mg/dm3, i.e. two to three orders of magnitude as less as in the investigated compositions. The discovery of the influence of alkali metal hydroxides on the formation of hydrogen with the participation of structural materials based on the example of aluminum makes it possible to suggest that the hydroxides of these metals contained in the coolant in a small amount can also take part in the hydroheterogeneous process of formation of minor hydrogen amounts. The potential for hydrogen formation in such a way needs to be taken into account during long-term operation of nuclear reactors, and during accidents and incidents at NPPs

Continuous online monitoring in pressurized water reactors during flexible operation using PLSR-based technique – Case study: Load following test

This paper proposes the application of a multivariant-based technique, namely, Partial Least Square Regression (PLSR) to enhance the process of continuous online monitoring and control in pressurized water reactors (PWR) during flexible operation. The results of the investigations of this paper have shown that the proposed technique provided effective performance during flexible PWR operation even with a small size of available history operating data samples. More specific, the proposed model provides the reactor operator with a comprehensive and simple presentation of the reactor status via a single dynamic trajectory during flexible operation which helps detecting anomalies at an early stage. In addition, results also have shown that the proposed PLSR-based model minimized the limitation on the ex-core accuracy resulting from the operation in the flexible mode, provided reliable, fast with excellent monitoring efficiency. Moreover, the results have also shown that the proposed model provides a graphical control guide which is achieved via a prioritized, reduced, comprehensive and manageable set of the PWR physical parameters. The analysis of this graphical guide has shown that it provides the fine control tuning, which in many cases is left for the operator, thus it minimizes the human error by relieving operators from having to pay attention to additional tasks during transient contingencies. In the context of this paper, a load-following test is conducted to ascertain the efficiency of the proposed technique. The data of the test was observed at Unit-1, Tihange nuclear power station, Belgium (three-loop PWR, 2873 MWth, 962 MWe net). The PLSR-based model for continuous model for continuous monitoring as well as the graphical control guide are developed and validated using SIMCA software package.

Study Protocol for Radiation Exposure and Cancer Risk Assessment: The Taiwan Nuclear Power Plants and Epidemiology Cohort Study (TNPECS).

This cohort was established to evaluate whether 38-year radiation exposure (since the start of nuclear reactor operations) is related to cancer risk in residents near three nuclear power plants (NPPs). This cohort study enrolled all residents who lived within 8 km of any of the three NPPs in Taiwan from 1978 to 2016 (n = 214,502; person-years = 4,660,189). The control population (n = 257,475; person-years = 6,282,390) from three towns comprised all residents having lived more than 15 km from all three NPPs. Radiation exposure will be assessed via computer programs GASPAR-II and LADTAP-II by following methodologies provided in the United States Nuclear Regulatory Commission regulatory guides. We calculated the cumulative individual tissue organ equivalent dose and cumulative effective dose for each resident. This study presents the number of new cancer cases and prevalence in the residence-nearest NPP group and control group in the 38-year research observation period. TNPECS provides a valuable platform for research and opens unique possibilities for testing whether radiation exposure since the start of operations of nuclear reactors will affect health across the life course. The release of radioactive nuclear species caused by the operation of NPPs caused residents to have an effective dose between 10-7 and 10-3 mSv/year. The mean cumulative medical radiation exposure dose between the residence-nearest NPP group and the control group was not different (7.69; standard deviation, 18.39 mSv and 7.61; standard deviation, 19.17 mSv; P = 0.114).

Two-group phase distribution characteristics for air-water flow in 5 × 5 vertical rod bundle channel with mixing vane spacer grids

The phase distribution in rod bundle channel was associated with the heat transfer coefficient, flow resistance and critical heat flux, which was of great significance for the design and safe operation of nuclear reactor and steam generator. In order to clarify the phase distribution characteristics and reveal the effect of mixing vane spacer grid (MVSG) in two-group air-water flow in rod bundle channel, the adiabatic upward air-water two-phase flow experiments were conducted in 5 × 5 rod bundles with MVSGs and the detailed local two-phase flow parameters were acquired with miniaturized four-sensor conductivity probe (MFSCP). Some interesting phenomena were reported: Group I and II void fraction profiles showed the core-peak features; Group I void fraction almost stayed at the region from 0.2 to 0.3 for two-group air-water flow at current flow conditions; the dissipation length of MVSG was larger than 32.1 L/D at current two-group flow conditions; and the influencing region of MVSG was divided into at least three regions along the axial direction with the flow development for current flow conditions.

Environmental impact assessments and long-term operation of nuclear power reactors: Increasing importance of environmental protection in the European Union?

Over the last decades, the world has become increasingly mindful of the potential impact of human activities on the environment and many governments have recognised the need to better assess the potential impacts of such activities prior to their authorisation. Legal requirements to perform environmental impact assessments (EIA) for activities likely to have environmental effects have been enacted under national, European and international law and have gained the status of international customary law.1 Due to the nature of the risks posed by the civilian use of nuclear energy on human health and the environment, it has become a requirement in most – if not all – countries that an EIA be performed prior to authorising the construction and operation of any nuclear installation, including nuclear power reactors.

The nuclear accident at Chernobyl: Immediate and further consequences

The accident at Chernobyl occurred in April 1986 at the Chernobyl Nuclear Power Plant in Soviet Union. The incident occurred during a scheduled safety test. A combination of inherent reactor design flaws and operators’ mistakes resulted in reactor’s No.4 disaster and the emission of a large quantity of radiation. The immediate actions involved the fire extinguishing, the cleanup of radioactive residues and the prevention of a new explosion. For this purpose, plenty of people worked with self-sacrifice. The people who lived nearby were removed. As far as the socio-economic impact for the Soviet Union is concerned, it was quite serious. Moreover, the environmental and human health consequences were also alarming with thyroid cancer being the most studied. Useful conclusions, especially for the safety both of reactors and nuclear power, as well as for the impact of radiation at ecosystems have been drawn. The debate about the use of nuclear power has remained open ever since.

Fuel rod nonlinear vibrations to detect and characterize Pellet-Cladding Interaction

Pellet-Cladding Interaction (PCI) in a Light Water Reactor is one of the major concerns to guarantee clad integrity while attempting at increasing the flexibility of PWR nuclear reactor operations to follow power grid demand. In order to forbid operations leading to clad failure, modeling capability to simulate the mechanism has improved through the years. Code development needs detailed and precise experimental data. Those data result from dedicated irradiation programs, named “power ramp tests”, carried out in experimental devices in Material Testing Reactors (e.g. in France ISABELLE in OSIRIS in the past (Alberman et al., 1997), the ADELINE loop in the Jules Horowitz Reactor JHR in the next future (Cheymol et al., 2011)). Those irradiation devices are highly instrumented to collect the most relevant information with the highest possible accuracy. In parallel with information gained thanks to post-irradiation examination programs, research is working on innovative methodologies to detect and characterize PCI kinetics during the tests. In this frame, we investigate the technological feasibility to measure the effects of PCI, on the vibrations of a nuclear fuel rod externally submitted to the turbulent axial flow rate excitation. We designed and realized the out-of-pile IMPIGRITIA set-up to reproduce, in controlled laboratory conditions, the mechanical interaction originating in the nuclear reactor. A single PWR rod test section, reproducing the main relevant geometrical and material characteristics of ADELINE experimental system, is submitted to the hydraulic excitation representative of the real case. The local closure of the gap is obtained by means of a remotely controlled expansion system. Measurements of the transverse displacement of the sample rod are collected by Laser Doppler Vibrometry. In this paper we introduce the design of the mock-up and the associated measurement method, then we present the two experimental phases and their results: the first one in air, to show the feasibility of the measurement in air in controlled conditions; the second one under turbulent flow rate to state on the feasibility of the passive detection. At the end, conclusions and perspectives for the work are discussed.