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The integration of nuclear reactors into the smart electric grid dictates the digitalization of Instrumentation and Control (I&C) systems, to support advanced networking capabilities (e.g., remote, semi-autonomous operation). However, continuous data transmission could potentially facilitate cyber events targeting nuclear installations. To secure nuclear communications against potential adversaries attempting to impersonate legitimate communicating parties, we propose a real-time processing scheme which does not rely on AI/ML, but uses noise analysis to verify the integrity of exchanged signals and commands. In this work, we evaluate the randomness of noise present in real-world nuclear I&C signals. After constructing a dataset consisting of steady state signals from PUR-1 all-digital reactor, three representative signals are used as case study. The disentropy of the autocorrelation function is used to quantify the amount of randomness present, while the signal observation time window is varied to determine the optimal monitoring length in a real-time cyber event detection scenario. We observe that several I&C signals contain additive components qualifying as white noise, which can therefore be used to evaluate their integrity. By performing several tests replicating cyber event scenarios, using original and adversary-manipulated data, we confirm the validity of our scheme for detecting False Data Injection (FDI) events.

Hardware-in-the-loop systems are a type of Hard real-time system that require a faster than real-time simulation. A method of running hard real-time systems with slower than real-time simulations is devised. Instead of waiting for the real-time subsystem output to a simulation, multiple simulation input predictions are generated and used to run several parallel simulation paths independent of the real-time subsystem. The simulation path results are stored until needed by the real-time subsystem at which time one is selected and returned to it to continue the execution of the system. A test case using a small modular reactor nuclear power plant hardware-in-the-loop system was used to demonstrate the effects of slower than real-time simulations on a hard real-time system and to show that this method allows such a simulation to be used. [ABSTRACT FROM AUTHOR]

This article aims at developing a multigene symbolic regression model of interfacial tension (IFT) for determining the health index of insulating oil. The oil parameters considered for modeling of IFT are break down voltage, resistivity, acidity, tan delta and water. The unique study undertaken in this article attempts to develop a multigene symbolic regression model for generating and evaluating the relationship between IFT and other oil parameters. Data utilized in this study involves actual oil testing parameters of in-service transformers and reactors of age more than 15 years. A sensitivity analysis has been carried out to investigate the relative contribution of input variables to the model output. The result of sensitivity analysis has further been used to determine the weight of parameters for calculating the health index of insulating oil. Finally, the health index of 400 in-service transformer and reactors are evaluated and analyzed. This research will help in undertaking decisions for the replacement of oil based on health index, thereby increasing the life span of transformer and reactors. The health index so obtained will act as a reference for planning necessary maintenance activities of the equipment and thereby saving unprecedented failure of the oil-filled transformer and reactors. [ABSTRACT FROM AUTHOR]

In recent years, rare-earth permanent magnets, including Nd2Fe14B and Sm2Co17, which are used in environments with high radiation, such as in electric motors for space equipment and robots operating in nuclear reactors, have become increasingly important. However, the demagnetization of the rare-earth permanent magnets induced by exposure to radiation has become a serious issue. In particular, when high-energy neutrons collide with the atoms in the magnet, the atoms can be removed and point defects can be generated. The spin fluctuations caused by point defects result in demagnetization of the rare-earth permanent magnets. In this study, we investigate via first-principle calculations, the probability of generation of point defects and the change in the magnetic moment in Nd2Fe14B and Sm2Co17 that occurs as a result of the point defect creation. We find that the generation probability of the point defects is very similar in both rare-earth magnets. The attenuation of the magnetic moment in Sm2Co17 was found to be considerably smaller than that of Nd2Fe14B. Thus, we conclude that the radiation resistance of the magnetic properties of Sm2Co17 is higher than that of Nd2Fe14B. [ABSTRACT FROM AUTHOR]

Planning of biorefinery operations is complicated by the stochastic nature of physical and chemical characteristics of biomass feedstock, such as, moisture level and carbohydrate content. Biomass characteristics affect the performance of the equipment which feed the reactor and the efficiency of the conversion process in a biorefinery. We propose a stochastic optimization model to identify a blend of feedstocks, inventory levels, and operating conditions of equipment to ensure a continuous flowing of biomass to the reactor while meeting the requirements of the biochemical conversion process. We propose a sample average approximation (SAA) of the model, and develop an efficient algorithm to solve the SAA model. A feedstock preprocessing process consists of two-stage grinding and pelleting is used to develop a case study. Extensive numerical analysis are conducted which lead to a number of observations. Our main observation is that sequencing bales based on moisture level and carbohydrate content leads to robust solutions that improve processing time and processing rate of the reactor. We provide a number of managerial insights that facilitate the implementation of the model proposed. Note to Practitioners—This paper is motivated by the challenges faced in the bioenergy industry. The focus of this paper is on plants which use the biochemical conversion process to generate liquid fuels. It has been observed that variations in biomass characteristics, such as moisture content, cause variations in feeding of the system which lead to under-utilization of equipment. A requirement of biochemical conversion process is to maintain the carbohydrate content of biomass processed by the reactor, larger than a threshold. We propose a model that identifies the inventory levels and operating conditions of equipment to ensure a continuous flowing of biomass to the reactor. The goal is to improve equipment utilization while satisfying the requirements of the conversion process. The model is tested using real-life data. We found out that by sequencing bales based on moisture level and carbohydrate content, a plant can reduce variability in the system leading to improved system reliability, higher processing rates of the reactor, and higher throughput. [ABSTRACT FROM AUTHOR]

The port-Hamiltonian form (PHF) of nuclear reactor dynamics is first given with the corresponding Hamiltonian function determined by the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics. Then, a passivity-based control (PBC) is proposed by adding extra damping injections to the PHF, and it is shown by theoretical analysis that satisfactory closed-loop stability can be guaranteed. The newly designed control law is applied to a nuclear heating reactor, and the simulation results not only verify the theoretical analysis but also show the satisfactory reactor power-level control performance. [ABSTRACT FROM AUTHOR]

We present a novel technique of neutron noise detection and experimental data interpretation developed during the EU H2020 project CORTEX aiming to improve the capabilities for identification and localization of neutron noise sources. The experimental data analysis is performed in the frequency domain by extracting the spectral power density and the phase angle using a novel spectral variance reduction technique based on per cycle-based bootstrapping with replacement. This technique allows for variance reduction of measured spectral power and phase angle not only at base frequency, but at higher harmonic frequency contributions as well. This allows for a more representative treatment of experimental data and validation of codes for neutron noise propagation, some of which have been developed within the project. The detector response is not necessarily linearly dependent on the oscillator movement, and the study of nonlinear terms provides additional information which can improve the accuracy of neutron noise source identification and localization. Moreover, these terms can be used in a Taylor series to identify a more complex dependence. The process is simplified in the sense that these contributions are linearized in the frequency domain as higher harmonic frequency contributions and can be easily identified and extracted due to spectral peaks prominence provided by the bootstrapping with replacement method. Combined with the spectral power and phase, we present a preliminary investigation of usability of higher-order terms for noise source identification and localization. In this article, we outline the CORTEX project, the experiments, and the measurement analysis methodology based on the bootstrapping with replacement along with the initial developments on the study of nonlinear terms in CROCUS reactor, using several different noise source configurations at a set frequency. [ABSTRACT FROM AUTHOR]

This article reviews the work to date on the CALOrimeter with Radial thermal transfers for nuclear REactors (CALORRE) differential calorimeter patented by Aix-Marseille University (AMU) and the French Alternative Energies and Atomic Energy Commission (CEA) in 2015. The article presents the results obtained with the first prototype of the CALORRE calorimeter qualified under real conditions during an irradiation campaign in the MARIA reactor in 2015, including previously unpublished details. Then, studies of different CALORRE calorimetric cells characterized by experiments under laboratory conditions are described. Several configurations were studied to determine the influence of cell height, horizontal fin geometry, and structural material composition on calorimeter response. These calculations provide for a calibration protocol by generating a heat source inside each cell, with evaluation of linearity, sensitivity, range, reproducibility, response time, and absolute temperatures. Finally, within the framework of a new research program called Compact-CALORimeter Irradiations inside the MIT research reactor (CALOR-I) and financed by AMU Foundation (A*Midex), a design optimization of the calorimeter assembly was carried out in order to remove contact thermal resistances and provide a new very compact CALORRE calorimeter suited for the in-core water loop of the Massachusetts Institute of Technology (MIT) reactor (2 $\text{W}\cdot \text{g}^{-1}$ peak nuclear heating rate). The response of this new very compact calorimeter is estimated using 3-D numerical thermal simulations under real conditions. [ABSTRACT FROM AUTHOR]

To measure fuel rod swelling in a nuclear research reactor, our research has focused on the development of an optical confocal chromatic sensor, allowing contactless measurements of radial changes in fuel rods. This sensor must be able to operate in the harsh and hot environment of a reactor, especially under high neutron fluences up to $10^{19}~{n}_{\mathrm{ fast}}/\text {cm}^{2}$ and gamma doses of a few GGy. When exposed to such radiation levels, the properties of bulk optical glasses are known to be affected by two main phenomena: radiation-induced attenuation (RIA) and radiation-induced refractive index change (RIRIC). With the objective of testing glasses as candidates for a future confocal chromatic sensor, we created a dedicated setup for the online monitoring of the glasses’ RIRIC in the core reactor with a measurement principle relying on interferometry. Since not only the refractive index (RI) but also the length of the samples changes under neutron radiation through compaction, we measured the variation of the optical path (OP) ($nL$). To assess the performance of our setup, preliminary in-lab tests were performed to evaluate the feasibility of measuring the small RI changes caused by external temperature changes. Due to this study, we were able to obtain data that were not yet available in the literature regarding the temperature-induced RI variation of several bulk glasses, including radiation-hardened glasses, up to 350 °C. This information is crucial for the targeted application, as the confocal chromatic sensor will have to operate at such elevated temperatures. [ABSTRACT FROM AUTHOR]

Silicon carbide (SiC) semiconductor is an ideal material for solid-state nuclear radiation detectors to be used in high-temperature, high-radiation environments. Such harsh environments are typically encountered in nuclear reactor measurement locations as well as high-level radioactive waste and/or “hot” dismantling–decommissioning operations. In the present fleet of commercial nuclear reactors, temperatures in excess of 300 °C are often encountered, and temperatures up to 800 °C are anticipated in advanced reactor designs. The wide bandgap for SiC (3.27 eV) compared to more widely used semiconductors such as silicon (1.12 eV at room temperature) has allowed low-noise measurements to be carried out at temperatures up to 700 °C. The concentration of thermally induced charge carriers in SiC at 700 °C is about four orders of magnitude less than that of silicon at room temperature. Furthermore, SiC radiation detectors have been demonstrated to be much more resistant to the effects of radiation-induced damage than more conventional semiconductors, such as silicon, germanium, or cadmium zinc telluride (CZT), and have been demonstrated to be operational after extremely high gamma-ray, neutron, and charged-particle doses. The purpose of the present review is to provide an updated state of the art for SiC neutron detectors and to explore their applications in harsh high-temperature, high-radiation nuclear reactor applications. Conclusions related to the current state-of-the-art of SiC neutron detectors will be presented, and specific ideal applications will be discussed. [ABSTRACT FROM AUTHOR]

Small fission chambers manufactured by Photonis, Reuter-Stokes, CEA, and Centronic are routinely calibrated in standard neutron fields of the BR1 reactor at SCK CEN in Mol, Belgium. Two irradiation fields are available: fast spectrum in the MARK3 convertor and thermal spectrum in the Empty Cavity. In this work, for the first time the calibration procedure of larger fission chambers (Photonis CFUL01 and CFUM21 type) with the deposit length exceeding the length of the convertor is presented. Spatial energy correction factors taking into account the neutron flux shape beyond the convertor and radial gradient of the neutron flux were calculated with MCNP and experimentally validated. The combination of calibration (i.e., effective mass measurement) in fast and thermal irradiation fields allows for determination of the purity of 238U deposit. This is demonstrated on three CFUL01-type fission chambers with purity between 99.8% and 99.998%. [ABSTRACT FROM AUTHOR]

Nuclear reactor inspections are critical to ensure the safety and reliability of a nuclear facility’s operation. In Canada, ultrasonic testing (UT) is used to inspect the health of pressure tubes that are part of Canada’s Deuterium Uranium (CANDU) reactor’s fuel channels. Currently, analysis of UT scans is performed by manual visualization and measurement to locate, characterize, and disposition flaws. Therefore, there is motivation to develop an automated method that is fast and accurate. In this article, a proof of concept (PoC) that automates the detection of flaws in nuclear fuel channel UT scans using a convolutional neural network (CNN) is presented. The CNN model was trained after constructing a dataset using historical UT scans and the corresponding inspection results. The requirement for this prototype was to identify the location of at least a portion of each flaw in UT scans while minimizing false positives (FPs). The proposed CNN model achieves this target by automatically identifying at least a portion of each flaw where further manual analysis is performed to identify the width, the length, and the type of the flaw. [ABSTRACT FROM AUTHOR]

Understanding the core neutron distribution is an indispensable condition to ensure the safe operation and effective burn-up of reactor fuel. It has become one of the new ways for reactor monitoring to use ex-core monitoring signals to calculate the in-core information recently. In this article, a fully convolutional network (FCN)-based architecture was implemented, where the 2-D and 3-D fully convolutional neural networks were used to reconstruct the space- and energy-dependent 3-D neutron distribution of the reactor core through the off-situ measurement directly. The new method has been validated in the H. B. Robinson reactor. The differences between reconstruction results of neutron flux distribution and Monte Carlo (MC) calculation were mostly less than 2%. The average relative deviation of reconstructing the neutron density was kept at 5%. It was demonstrated that the provided method was applicable to monitor the reactor core theoretically and would provide strong support for the subsequent research on the real-time monitoring for an operating reactor, safety, and operational testing for new reactor designs, and evaluation of source term information in a nuclear emergency. [ABSTRACT FROM AUTHOR]

Fiber Bragg gratings (FBGs)-based temperature sensors present numerous advantages such as small packaging, fast acquisition rate, and accuracy for structural health-monitoring applications in nuclear environments. Among the various classes of FBGs, Type I FBGs are inscribed with a UV continuous or pulsed laser on a photosensitive fiber or with femtosecond laser even in nonphotosensitive fibers. These gratings, however, cannot survive at temperatures exceeding 400 °C. Regenerated FBG (RFBG) gratings, instead, are derivative from type I grating: when a high thermal treatment (> 650 °C) is applied after the inscription on a prehydrogenated fiber, a new grating (RFBG) appears with different thermal properties. It withstands temperatures as high as 1000 °C, opening the way to new application fields such as the temperature monitoring of the nuclear reactor cores. This work investigates the radiation response of RFBGs originated from type I seed FBGs inscribed with an argon laser (244 nm) in two different fibers (SMF-28e fiber or in a B/Ge co-doped fiber), loaded with either H2 or D2 before the inscription to enhance the fiber photosensitivity. Regeneration was achieved at 650 °C or 900 °C depending on the fiber type. After this, the RFBGs were irradiated under 40-keV X-rays at two different dose rates [1 Gy(SiO2)/s or 10 Gy/s] and at two temperatures of irradiation—25 °C or 250 °C. At room-temperature (RT) irradiation, a Bragg Wavelength Shift (BWS) of about 35 pm was observed for the SMF-28e and more than 130 pm for the B/Ge fiber at 400-kGy dose, and the combined temperature and radiation constraints reveal that the RFBGs are radiation-tolerant. Indeed, no BWS is observed anymore, at the highest temperature of irradiation. [ABSTRACT FROM AUTHOR]

A small modular reactor (SMR) offers a promising means to produce carbon-free power. Many SMR designs are intended for multiple module operation, whereby several units are operated as a single plant. As such, the output from the plant can be increased or decreased via regulating the individual units. In order to benefit from such operational flexibility, individual units must be properly controlled, as there are direct and indirect couplings among their dynamics. This research examines such a situation and analyzes the dynamic behavior of a group of pressurized water reactor-type SMRs in an islanded power system under the load-frequency control framework. The detailed models of the SMRs and the associated components and subsystems are developed, and effective control systems are designed to investigate the performance of this multimodule topology in supplying uninterrupted power to the loads under different operating scenarios. The performance of the system is evaluated through simulations in PSCAD/EMTDC using a benchmark 9-bus power system. [ABSTRACT FROM AUTHOR]

A proportional–integral extended state observer (PI-ESO) of nuclear reactors is proposed for disturbance estimation, driven by the measurement errors of neutron flux, primary coolant temperatures at reactor inlet and outlet, as well as their integrations. This PI-ESO can provide a globally asymptotically bounded estimation for not only the reactor process variables, including normalized neutron flux and precursor concentrations as well as average temperatures of fuel assemblies and primary coolant, but also the total disturbances in channels of neutron kinetics and coolant thermal hydraulics. Furthermore, the newly built PI-ESO is applied to reconstruct both the unmeasurable state variables and the total disturbances of a nuclear heating reactor (NHR). Numerical simulation results verify the theoretical result, showing both the satisfactory performance and the influence of observer parameters. [ABSTRACT FROM AUTHOR]

The superconducting induction pump is a pumping system which moves electro-conductive liquid by using electromagnetism. It has the potential to be applied as the power system for the submarine propulsion and the pumping system for nuclear reactors. Since alternating current may generate more heat in the superconducting winding, it is extremely essential to analyze AC loss characteristics of the pumping system to avoid quenching phenomenon of superconducting tapes. In this article, a 2D-axisymmetric model of an annular linear induction pump which is composed of superconducting coils has been established by using T-A formulation. AC loss of superconducting coils under self-field and external alternating magnetic field have been studied. Meanwhile, the influence of some essential parameters such as the transmission current, coil turns have been analyzed as well. Results show that the arrangement position of coils in the pump plays an important role in determining the the AC loss of the superconducting winding. Meanwhile, self-field of the superconducting coils seems to have larger influence on the AC loss characteristics than the external alternating field. In conclusion, mechanism and relevant action laws of AC loss characteristics in the superconducting linear induction pump transported with three-phase alternating current and under the condition of external background AC field have been found in this article. [ABSTRACT FROM AUTHOR]

Nuclear reactors are designed to maintain a chain reaction producing a steady flow of neutrons generated by fission of heavy nuclei, such as uranium. They produce thermal energy, which is converted into mechanical energy and ultimately into electrical energy feeding to a grid network. In a nuclear reactor, fuel undergoes various power/temperature transients during normal and off-normal situations. One needs to simulate these transients in experimental facilities for performing variety of reactor engineering and safety related tasks. The facility consists of scaled thermal hydraulic test setup comprising process equipment, instrumented nuclear fuel channel simulator, and high current converters for feeding power equivalent to fission power to those simulators to handle the transients safely in controlled manner. The experimental results are used to generate database of various vital nuclear reactor process parameters under all types of operational transients and off-normal conditions that can be compared with predictions from computer codes thus validating ever-evolving and advanced software. Besides, simulation of severe accidental conditions of nuclear reactor in facility is also necessary for evaluation of thermal hydraulic design safety margin. A low-resistance fuel simulator of identical geometry with power scaling is directly heated by high dc current to simulate fission power for generating experimental database. The converters have capability to generate power/temperature transients for simulating nuclear reactor conditions. This article presents design, functional description, specifications, simulation, hardware implementation of 10 kA, 12-pulse, thyristorized converters powering fuel simulators. The successful implementation of parallel operation of four converters feeding 27 kA current is described in this article. [ABSTRACT FROM AUTHOR]

High-energy (>40 MeV) protons are commonly used to characterize radiation tolerance of space electronics against damage caused by energy transfer to the nuclei and electrons of semiconductor materials while in orbit. While practically useful, these experiments are unrepresentative in terms of particle type and energy spectra, which results in disproportionate amounts of displacement damage and total ionizing dose. We compare these damages to those realized by bulk semiconductors used in optoelectronics in common low, medium, and high Earth orbits by calculating the duration in orbit required to achieve equivalent nuclear and electronic energy deposition. We conduct this analysis as a function of test proton energy, material, material thickness, and shielding thickness. The ratio of nuclear to electronic orbit duration, a value which would approach unity in an ideal radiation tolerance test, is found to exceed unity in the majority of cases but approaches unity as Al shielding increases. This study provides a connection between damage produced in terrestrial accelerator-based characterizations and orbit irradiation in terms of both damage modes which can cause optoelectronic components to fail: displacement damage and total ionizing dose. [ABSTRACT FROM AUTHOR]

Radiation-induced attenuation (RIA) at a wavelength of λ = 1.55 μm as well as the RIA spectra were investigated in optical fibers during and after irradiation at the IVG.1M nuclear reactor (Republic of Kazakhstan). The fibers studied were single mode at λ = 1.55μm, had an undoped silica core and an-F-doped silica cladding [pure-silica fibers (PSFs)], and were provided by different manufacturers. In addition to five PSFs, a single-mode revolver hollow-core fiber was also studied. The PSFs had different coating materials: polyimide, copper, or aluminum. All fibers were exposed to three runs of reactor irradiation at enhanced temperature (155–355 °C) so that by the end of the third run, the fast-neutron fluence amounted to Φ = 4.46 ⋅ 1017 n/cm2 and γ-dose to D = 2.91 ⋅107 Gy. The total induced loss at λ = 1.55μm was found to be composed of 1) RIA, 2) bending loss (BL), caused by thermal expansion of the copper cylinder the PSFs were wound on, and 3) thermo-induced microbending loss (ML) in Cu-coated PSFs. At the end of the irradiation campaign, the net RIA in PSFs proved to lie in the interval of 0.12–0.16 dB/m, which appears to be admissible to intrareactor transport fibers. It was confirmed that long-wavelength RIA tail is the only RIA origin at λ = 1.55μm in PSFs in such strong radiation fields. The hollow-core fiber was found to feature zeroth RIA throughout the campaign. [ABSTRACT FROM AUTHOR]

In this article, we propose a methodology based on cubature Kalman filtering (CKF) for simultaneous estimation of two important variables of nuclear reactors, viz. the neutron flux and the total core reactivity, from signals of delayed response self-powered neutron detectors (SPNDs). Moreover, by using the linearity of our model, the Rao-Blackwellized CKF (RBCKF) algorithm is developed for our estimation problem, which offers benefits in terms of smaller computational load in comparison to standard CKF. Furthermore, a scheme which is termed as adaptive-RBCKF (A-RBCKF) has been designed for adaptation of process noise covariance. A comparative study of the two state estimators, that is, RBCKF and A-RBCKF, through simulations under different transient scenarios demonstrates the efficacy of A-RBCKF over RBCKF. [ABSTRACT FROM AUTHOR]

Large nuclear reactors exhibit spatial oscillations in neutron flux distribution on account of xenon reactivity feedback. These oscillations, if not suppressed, can pose nonuniform spatial distribution of power and raise chances of fuel failure. Thus, it is necessary to design a control strategy to regulate these spatial oscillations. This article presents the design of an integral sliding mode control (ISMC) for spatial power control of the advanced heavy water reactor (AHWR). The AHWR model considered here has 18 outputs as well as 4 inputs and is characterized by 90 state variables. This nonlinear model is linearized around an operating point and an optimal controller is designed. To ensure the robustness of the system under external disturbances, ISMC is designed. The finite-time convergence of the controller is provided by the Lyapunov approach. A comparative simulation study is presented to show the efficacy and robustness of the proposed controller using the nonlinear model of AHWR. [ABSTRACT FROM AUTHOR]

Small modular reactors (SMRs), an emerging nuclear power plant technology, are suitable for large grids as well as remote load centers and offer load following and frequency response capabilities. While the SMRs have expectedly higher response rates, detailed dynamic models including reactor dynamics are necessary for power system dynamic studies. This paper presents a dynamic model of an integral pressurized water reactor (iPWR)-type SMR, modeled in Siemens PTI PSS/E, to assess the contribution of the reactor to the power system dynamics. The proposed SMR model mimics the heat generation process and subsequent heat transfer process with the inclusion of the reactor core based on point kinetics, primary coolant based on natural circulation, and a simplified three lump representation of the steam generator. Controllers are designed to operate the turbine valve and reactor control rod in closed loops. The SMR model is integrated with the modified turbine-governor system and a power system study is conducted. Results show the power system and internal reactor responses when subjected to electrical demand variations of 20 ${\%}$ rated electrical output (REO) with a valve rate limit of $\pm \text{80}{\%}$ REO/min. [ABSTRACT FROM AUTHOR]

This article considers the design of an optimal state feedback controller for a nuclear reactor. The model of 700-MWe Indian pressurized heavy water-type reactor is described and is considered for design. The cost function is the norm of the state feedback vector, which is minimized by making use of extra degrees of freedom obtained due to nonspecific pole placement in the left half of the s plane. The stability region is approximated by a maximal sphere in the co-efficient space of the closed-loop characteristic polynomial. The optimization problem is formulated as a quadratically constrained quadratic program and is solved using a primal-dual search method. The issues related to the choice of the center of the approximated sphere and correct step size in the search method are addressed. The norm of the state feedback vector obtained by this method is about 100 times less than those obtained using comparable methods and yet yields acceptable transient performance at various operating points of the reactor. [ABSTRACT FROM AUTHOR]

Xenon (Xe) is a noble gas with an isotope (Xe-135) that has an extremely high absorption cross section of neutrons moving at thermal energies. Xe-135 is produced through a chain of decay of isotopes resulting from nuclear fission, in addition to directly resulting from fission. The decay chain results in a phase lag between Xe-135 concentration and the neutron flux or power in a nuclear reactor. The phase lag results in oscillations and several hours of delay between the set power and the response power of a reactor. To enable rapid load following of reactors to a maximum of one change per hour, or a frequency of $2.7\times 10^{-4}$ Hz ($1.74\times 10^{-3}$ rad/s), a Bode-step controller is designed to expand the functional bandwidth of the reactor and to stabilize the control loop. The resulting controller is capable of projecting the Xe-135 concentrations and providing a phase lead to overcome the inherited phase lag. The controller was found to eliminate the power oscillation and enable the targeted “one change per hour” power set point change. [ABSTRACT FROM AUTHOR]

Core power control of a nuclear reactor during power maneuvering transients is an important issue in a nuclear power plant. This paper, for the first time, presents an optimized PID controller for a nuclear research reactor with a new model of the reactor core, which is based on the fractional neutron point kinetics (FNPK) equations. This model preserves the main dynamic properties of the neutron movement in which the relaxation time related to a quick change in the neutron flux includes a fractional order, acting as an exponent of the relaxation time, to show the dynamic characteristics of the nuclear reactor in the best way. The physical justification of the fractional exponent is associated with nonFickian impressions from the neutron diffusion equation point of view. In addition, in this paper, for the first time, feedback reactivity effects are considered for FNPK model. Genetic algorithm (GA) is used to optimize gains of the PID controller. The objective function is according to the minimization of the integral of time-weighted absolute error (ITAE). The simulation result demonstrates that this optimized control method has satisfactory performance and stability during output-tracking process for higher values of fractional order. [ABSTRACT FROM AUTHOR]

Late one spring night in 1986, around 1:30 a.m., the residents of Pripyat, a Ukrainian city of 50,000 people at the northern tip of the Dnieper River, were shaken from sleep by a giant explosion originating in the nuclear reactor of the nearby power plant in Chernobyl. It was an event that would forever change the course of their lives, and that of human history. Hundreds of people died, both directly and indirectly from the explosion, which is now regarded as the worst nuclear power plant catastrophe in history. Today, the Chernobyl power plant and Pripyat are completely abandoned, unsafe due to nuclear contamination. Within a 19-mile radius in every direction of the plant, the landscape is considered unfit for human life for the next 50,000 years. And the cause? Well, when the dust settled, it was a simple condition we all suffer from time to time: an error caused by a worker with too little sleep. [ABSTRACT FROM AUTHOR]

If the axial neutron flux distribution could not be controlled, the power at some locations in the nuclear reactor core might exceed the thermal limits, resulting in increased chances of fuel failure. In fact, the power is proportional to the neutron flux, it is necessary to develop a neutron flux distribution control method. However, the nuclear reactor is a kind of distributed parameter system (DPS) which brings a great challenge for the traditional lumped parameter control method. In this paper, we mainly focus on the fast reactor that present a distributed parameter control (DPC) method framework such that the closed-loop neutron flux distribution control system is exponentially stable. Initially, the closed-loop exponentially stable sufficient conditions of the control law for the neutron flux DPC problems are provided based on the integral Lyapunov function. The most important characteristic of the control law is that the absorption cross section needs to be changed globally and simultaneously in the axial direction. The analysis shows that the control law can be realized by the rotating drum theoretically because it can strictly conform to the reactivity insertion constraint of the control law. The spatial differential linear matrix inequality optimization techniques are employed to approximately solve the suboptimal integral control law design problem. The numerical experiment of the Lead-Bismuth Eutectic-cooled fast reactor shows that the initial neutron flux peaks could be inhibited by this distributed control method effectively and rapidly. [ABSTRACT FROM AUTHOR]

Terrestrial, nuclear reactor, and monoenergetic neutron-induced single-event upset (SEU) in SRAMs with different technology nodes more than $1.50~\mu \text{m}$ –40 nm were explored experimentally. Monte Carlo simulation was accomplished in Geant4 software to calculate neutron-induced SEU cross section and energy deposition of neutron-induced secondary particles in SRAMs. The results indicate that the neutron-induced SEU cross section in SRAMs changes in a nonmonotonic way with the scaling down of technology nodes. For 0.5–0.13- $\mu \text{m}$ SRAMs from HITACHI series, SEU cross section increases with the decrease of technology node, both in terrestrial and nuclear reactor neutrons. For $0.13~\mu \text{m}$ –40 nm SRAMs from ISIS series, SEU cross section decreases with the decrease of technology node, and this trend keeps the same in 2.5 MeV, 14 MeV, and reactor neutron irradiation, respectively. The threshold energy of neutron-induced SEU decreases sharply with the decreasing technology nodes, and 16 O, 28 Si, $\alpha $ particles, and 24 Mg are the main secondary particles inducing SEU. [ABSTRACT FROM AUTHOR]

In this paper, one proposes a model-assisted probability of detection approach to study the reliability of detecting special defects. The approach is based on a 3-D finite-element T– $\Omega $ formulation with circuit coupling. This model is then applied to calculate performances of induction thermography method to detect thin open cracks in vessel reactor shell in nuclear installation. [ABSTRACT FROM AUTHOR]

In situ current–voltage characteristics of Nb-doped TiO2/p-Si-based heterojunction diode have been studied under dense electronic excitations of 84-MeV Si6+ ions. The diode parameters such as ideality factor ($\eta $), barrier height ($\phi _{B}$), reverse saturation current (${J}_{s}$), and series resistance (${R}_{s}$) are found to be a strong function of ion irradiation fluence. The observed anomalies of fluence dependence of barrier height and ideality factor are explained in terms of irradiation-induced created defects complexes, modification of interface states and structural properties of Nb-doped TiO2 (NTO) layer. Several mechanisms such as barrier height inhomogeneity, donor defects-induced enchantment in n-NTO layer conductivity, and various current conduction mechanisms involved at different voltage ranges are discussed as a function of fluence with the help of constructed energy band diagram. Such in situ studies on n-NTO/p-Si heterojunction diode under radiation harsh environment are very appropriate for the better understanding of heterojunction interface properties and make it suitable for use in aerospace industry and nuclear reactors. [ABSTRACT FROM AUTHOR]

Reactivity is widely used as the paramount means for defining nuclear reactor status, but there is no physical sensor available to measure this parameter directly. Similarly delayed neutron precursors’ concentrations, the source of the delayed neutrons which play important role in reactor control cannot be measured directly. Hence, these quantities must be calculated or estimated from observed neutron flux transients. Recently, Kalman filter (KF) and its variants have been used for the reactivity estimation. A key limitation of KF and all of its variants is that they cannot account for the constraints on state variables systematically, which may cause the filter’s instability. In this paper, a constrained estimator based on the recursive nonlinear dynamic data reconciliation technique is proposed for the estimation of reactivity and delayed neutron precursors’ concentrations in the nuclear reactor. The effectiveness of the proposed technique has been evaluated through the step test signal in the presence of measurement noise and experimental power variation data sets collected from one of the Indian reactors. [ABSTRACT FROM AUTHOR]

Traditional heavy-ion testing for single-event effects is carried out in cyclotron facilities with energies around 10 MeV/n. Despite their capability of providing a broad range of linear energy transfer (LET) values, the main limitations are related to the need of testing in a vacuum and with the sensitive region of the components accessible to the low range ions. In this paper, we explore the use of ultrahigh energy (UHE) (5–150 GeV/n) ions in the CERN accelerator complex for radiation effects on electronics testing. At these energies, we show, both through simulations and experimental data, the significant impact of the ion energy on the ionization track structure and associated volume-restricted LET value, highlighting the possible limitations for radiation hardness assurance for high-energy accelerator applications. In addition, we show that from a nuclear interaction perspective, UHE ions behave similar to protons independently of their significantly larger mass. [ABSTRACT FROM AUTHOR]

Inferring the ON/OFF operational state of a reactor facility using measurements from an independent monitoring system is critical to the assessment of its compliance to agreements. We consider the problem of inferring the ON/OFF state of a reactor facility using the measurements of Ar-41, Cs-138, and Xe-138 gas effluence types collected at the facility’s off-gas stack. We present classifiers based on thresholding measurements of individual effluence types, and then present fusers that combine their outputs or measurements. We present five fusers based on the simple majority rule, Chow’s pattern recognition function, Fisher’s combined ${p}$ -value statistic, the physics-based Poisson radiation counts model, and the correlation coefficient (CC) method. In addition, we also test five machine learning methods based on nonlinear classifiers, which are available as R packages. Our results show that: 1) these gas effluence measurements are effective in inferring the ON/OFF state of a reactor facility, for example, best fusers achieve ~97% detection at ~1% false alarm rate and 2) fusers that combine all effluence types based on physics-based models, CC, and Fisher’s method outperform the simple majority rule, Chow’s fusers, and the machine learning methods, as well as when they are applied to individual and pairs of effluence types. [ABSTRACT FROM AUTHOR]

This paper presents the results of the fission chambers (FC) behavior under intensive neutron and gamma radiation at the research reactors. The FC detector unit (DU) was designed for the International Experimental Thermonuclear Reactor neutron diagnostic “divertor neutron flux monitor.” Two mockups of the DU with FCs containing 1 g and 100 mg of 235U were tested. The test complex is based on the FC signal processing unit and the FC parameters’ monitoring system. The processing unit contains several measuring chains and acquire FC signal over a wide range of neutron flux variation. The FC parameters’ monitoring system records signal shapes, acquires pulse-height amplitude spectra, and checks electrical parameters of the neutron detector. The radiation tests were carried out at the research reactors IRT MEPhI and VVR-c. The structure and characteristics of the test complex as well measurement techniques and tools for FC characterization are discussed. The results of reactor experiments to determine characteristics of the DU mockup and wide-range neutron flux monitoring channels are presented. [ABSTRACT FROM AUTHOR]

In this paper, test measurements of a fibered optically stimulated luminescent dosimeter (OSLD) system performed during a dedicated experimental phase in EOLE zero-power reactor are presented. The measurement setup consists of an OSLD crystal connected onto the extremity of an optical fiber and a laser stimulation system, manufactured by the CEA/LIST, Saclay, France. The OSL sensor is remotely stimulated via an optical fiber using a diode-pumped solid-state laser. The OSL light is collected and guided back along the same fiber to a photomultiplier tube through appropriate optical filters. OSL signals are corrected for background noise and integrated to provide the absorbed doses thanks to calibration coefficients. The light stimulation also fully bleaches the crystal (reusable for the next measurement). Results obtained using this system are compared to usual gamma-heating measurement protocol using OSLD pellets. Both measurement processes were calibrated to ensure the comparison of absolute doses in gray at the same positions in the core. The presence of induced radioluminescence in the OSLD during the irradiation was also observed and could be used to monitor the gamma flux. The feasibility of remote measurements is achieved, whereas further developments could be conducted to improve this technique since the readout procedure still requires to withdraw the OSLD off the gamma flux (hence from the core) on account of the dose rate (around a few Gy $\cdot \text{h}^{\mathrm {\mathbf {-1}}}$), and the readout time remains quite long for online applications. Several improvements are foreseen and will be tested in the forthcoming years. [ABSTRACT FROM AUTHOR]

The central solenoid (CS) model coil (CSMC), which has a high stored energy, is developed to validate the CS manufacturing methods for the China Fusion Engineering Test Reactor (CFETR). The designed operation current of the model coil is 47.65 kA. The maximum rate of the magnetic field is 1.5 T/s. The stored energy of the model coil is 407.6 MJ. A set of verification test of the model coil will be conducted after manufacture, so a CSMC power supply, which generates the current for coil operation, and a quench protection, which could quickly release the stored energy, are required. The main circuit topology of power supply was designed according to the requirements of the testing current waveform. The current control was adopted, which controlled the actual output current provided by the power supply system to track any specified current reference of the model coil within acceptable precision. The design of the quench protection system for model coil was also described in this paper. [ABSTRACT FROM AUTHOR]

This paper describes the high-priority testing activities supporting the ITER radial neutron camera (RNC) design, performed by a consortium of European institutes within a framework contract placed by fusion for energy, the ITER European Domestic Agency. The main role of the RNC is to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium and deuterium–deuterium fusion reactions through an array of flux monitors/spectrometers located in collimated lines of sight viewing the plasma through the ITER equatorial port plug #1. The line-integrated neutron fluxes will be used to evaluate, through reconstruction techniques, the radial profile of the neutrons emitted per unit time and volume (neutron emissivity) and, therefore, the neutron yield and the alpha particles’ birth profile. The activity of high-priority testing is dedicated to the preparation and the design of experimental test environment, the conduction of appropriate tests and reporting of test results for the high-priority prototypes, clarifying or verifying the expected key function and system behavior, and enhancing learning on specific issues (potential showstoppers). [ABSTRACT FROM AUTHOR]

Vacuum vessel (VV) is the main component of the China fusion engineering test reactor. Shell forming is relatively in the earliest phase, one of the most important processes during VV manufacturing. The profile errors caused by spring back of shells at room temperature will directly influence the subsequent manufacturing and experimental studies, so a set of dies, whose profiles can be easily revised, has been designed based on the spring-back theory. The empirical formula has been used to calculate the theoretic die profile value, which is mostly determined by the mechanical properties of 316LN stainless steel. Then, choosing four relevant values to simulate the forming process in which the optimal profile errors of shell must be controlled within 1.5 mm. The optimal dimensional value of the dies has been confirmed to guide the forming experimental study. After the experiment, the actual profile errors, thickness reduction, maximum spring back, maximum deformation rate, and the surface residual stress have been measured, and those data are an accurate indicator from the finite element analysis, also verifying the reasonability of the manufacturing process in turn. [ABSTRACT FROM AUTHOR]

A successful conceptual design was completed to develop in-vessel control coils [non-axisymmetric control coils (NCCs)]. The NCCs are a series of saddle coils that are intended to satisfy a number of physics criteria including magnetic breaking, error field control, fast resistive wall mode (RWM) control, and edge-localized-mode stabilization. Customized mineral-insulated cable (MIC) was selected for the conductor material. The MIC was made from oxygen-free copper conductor, high purity magnesium oxide powder insulation, and 304 stainless steel sleeve cover. Sample MIC was purchased from an outside supplier, and various tests were conducted for design and performance verification including high-voltage testing, which necessitated the development of special test terminations. A concept design was also developed for terminating the MIC ends using the nonconductive and vacuum-sealed technique. Two alternative designs were proposed for joints inside the vacuum vessel (VV). The NCCs are designed to be mounted in front of the primary passive plates and underneath the plasma-facing component (PFC) tiles. The passive plates will be modified to accommodate the coils. A new PFC tile design concept was developed using high-Z materials. New designs for the penetrations through the VV wall were developed. One port per coil was planned. A new power patch panel will be required to provide the ability for various combinations of connections between the NCC, the existing RWM coils, and the existing switching power amplifier power suppliers. [ABSTRACT FROM AUTHOR]

In order to develop and verify the key manufacturing technologies of the large Nb3Sn superconducting magnet, the central solenoid model coil (CSMC) for China Fusion Engineering Test Reactor is being developed in the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, China. The design value of the highest magnetic field for CSMC is 12 T, which is made of Nb3Sn and NbTi hybrid superconducting magnets. The key structural parameters of all windings have been determined by the electromagnetic design and optimization of the coil. All the windings of CSMC are a kind of pancake coil wound one in hand by cable-in-conduit conductor; each winding is composed of pancakes (concentric circular turns with turn transitions), pancake joggles, and upper and lower leads. The winding design has been finished including the winding direction analysis, the key bending segments geometry of turn transitions, pancake joggles, and leads transitions. Besides, the new winding production line has come into use and some key forming trials have been conducted. [ABSTRACT FROM AUTHOR]

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipes of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper. [ABSTRACT FROM AUTHOR]

Based on the finite-state machine (FSM) theory, a method of HL-2A host centralized control system FSM model for achieving centralized control and management is proposed. This method integrates all subcontrol systems and realizes interlock protection between subsystems. The tool of stateflow combined with Simulink is adopted to achieve the aided design and dynamic simulation of the FSM model. In-depth development of Experimental Physics and Industrial Control System (EPICS), sequencer toolkit is taken to realize the FSM model-specific operation. The result indicates that the FSM model can run on the EPICS platform successfully and establish significant basis for design of the next-generation device host centralized control system. [ABSTRACT FROM AUTHOR]

To achieve the maximum current rising rate, the second-generation fast control power supply (FCPS) applies voltage control mode while retaining the first-generation current tracking mode. When the power supply outputs high voltage, the accompanied frequent overcurrent protection will return the power to dc side, which leads to overvoltage protection, so the current limiting protection is applied. Experimental results prove that voltage mode can greatly improve the ability to suppress vertical unstable displacement of plasma. [ABSTRACT FROM AUTHOR]

In 2015, the optimized stellarator Wendelstein 7-X stellarator (W7-X) started with operation. The main objective of W7-X is the demonstration of the integrated reactor potential of the optimized stellarator line. An important element of this mission is the achievement of high heating power and high confinement in the steady-state operation. The approach to this mission is the following three steps. First, plasmas were produced in a limiter configuration [operation phase (OP 1.1)], then a test divertor unit is being installed (temporary divertor unit) for the next campaign, OP 1.2, before the full steady-state capability will be achieved implementing active cooling of all in-vessel components and a steady-state high heat-flux divertor. In December 2015, the first helium plasma was generated using electron cyclotron resonance heating (ECRH), in February 2016, the working gas was switched to hydrogen. The first OP (OP 1.1) was successfully finished in March 2016. At the end of OP 1.1, the discharge duration was close to 6 s, the limit for the integrated heating power was increased to 4 MJ and electron temperatures of ~10 keV were achieved. Due to the low densities in the range of 1019 cm−3 and the pure electron heating by ECRH, the ion temperatures reached only 2 keV. At present, W7-X is undergoing the next completion phase, including the installation of the test divertor unit, the installation of the carbon tiles on the inner plasma vessel wall, an upgrade of existing diagnostics, and the installation of new diagnostics. This paper discusses the first operational phase, lessons learned and implemented, and the status before the start of the second OP (OP 1.2). [ABSTRACT FROM AUTHOR]

The external bypass, as an important component of the international thermonuclear experimental reactor (ITER) poloidal field converter unit (PFCU), will provide a freewheeling loop for the load current to protect the magnets and PF converter modules from being damaged by overcurrent and overvoltage under fault conditions. The triggering bypass protection test is used to verify that the designed bypass is triggered normally and endure the rated load current. In this paper, a suitable test scheme is designed for triggering bypass protection test of ITER PFCU based on the PF converter integrated test platform at the Institute of Plasma Physics, Chinese Academy of Sciences. This test scheme includes the triggering bypass method, calculating the trigger angle of the converter in inverter and a method of reducing the bypass current to zero in the absence of ITER mechanical switch. The feasibility of the test scheme is successfully verified by the simulation results and test results, both of which prove that the external bypass of ITER PFCU is triggered normally and withstands the load current for 100 ms. Due to the integrated inductance of the actual dc output circuit, the actual transfer time of bypass current in the test is more than the simulation results. [ABSTRACT FROM AUTHOR]

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥1 km/s. An innovative approach, aimed at identifying suitable inboard “direct line” paths, to inject high-speed pellets (in the 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. A possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined. [ABSTRACT FROM AUTHOR]

Injection of solid nonfuel pellets has been actively used as a tool for pacing and mitigation of edge localized modes (ELMs). In DIII-D, effective ELM pacing has been demonstrated by the high-frequency injection of lithium and carbon submillimeter spheres, using the impurity granule injector (IGI). This device injects granules into the plasma at speeds up to 150 m/s, through impact with a rotating impeller. In the IGI, high-frequency granule delivery was accomplished through a vibrational granule dropper, in which high time-average rates are obtained at the cost of lack of period control. We present a new in-line granule feeder, capable of delivering granules of size 0.2–2 mm with no restriction of material properties, at quasi-periodic rates of up to 150 Hz, for 0.7-mm diameter lithium granules (600 Hz using 0.3-mm granules). The new dropper mechanism combines two piezo-in-line units; one which feeds the impeller and one which circulates granules that are filtered out of the feeder path. A remotely adjustable filter eliminates granules that are stacked, oversized, or side by side allowing the formation of a single moving granule injection line. The granules fall off the in-line feeder exit one at a time, achieving a quasi-periodic delivery rate proportional to the exit speed. At higher rates, the periodicity deteriorates. This behavior was studied using high-speed cameras and electrostatic measurements, and it was found that at drop rates <60 Hz, the granule delivery period has a variation of ±25% which appears to be caused by gaps which develop in the last centimeter of the injection line, as granules exit off the moving track. The linear feeder concept is robust against bridge instabilities and clogging issues, thanks to the simple diverter filter and constant recirculation of granules. Furthermore, the open-top design of the device facilitates refilling the device from separate reservoirs and has easy access for directly monitoring operation and adjustment. This paper describes the details of the in-line feeder design, along with several design iterations. The goal is a robust in-vacuum mechanism that can deliver granule flow ranging from a single particle to a line of particles at 150 per second, using different sizes and materials from the same apparatus. [ABSTRACT FROM AUTHOR]

Magnetic compression (MC) technology was recommended for tokamak to study compressed plasma and a power module have been designed here for developing its server power supplier. The minor-radius compression is one of the most effective methods to improve the performance parameters of existing tokamaks, enabling the fusion plasmas operated at high density, high temperature, and high beta. In this paper, a high-frequency and high-power supply system is proposed, used for powering coils of MC within the vacuum chamber of the experimental advanced superconducting tokamak. The basic of the power module is an ac/pulse converter, implemented by a series-resonant full-bridge circuit and controlled by pulsewidth modulated. Also, control method adopts current closed-loop proportional–integral control, which has less than 3-ms current response time in real time. The converter is analyzed and the design procedure is discussed. Experimental results obtained from a 3-kW converter prototype are presented to validate the converter’s performance with the redesigned control board. [ABSTRACT FROM AUTHOR]