Core Power Research Articles

Nuclear Reactor Power Level Model Predictive Control: A Consideration of Coolant Outlet Temperature Relaxation Tracking Method

Nuclear reactor core power level control and the average coolant temperature usually adopt independent control strategies, but tracking only one single core power makes the coolant temperature control risky in a disturbance situation. To solve this problem, a reference trajectory calculation based on the coolant outlet temperature under the steady state of the reactor power is designed; then, a model predictive controller with a dual objective optimization function, which also includes a relaxation factor, is constructed. By setting the tracking weight matrix in the controller, the tendency setting of dual target trajectory tracking can be realized. Simulation experiments verify the effectiveness of core power level control considering the coolant outlet temperature trajectory tracking. Furthermore, the tracking effect of different weight matrices is simulated and analyzed. The results show that the developed control strategy is more flexible than a single control scheme.

自由界面渦による気相巻き込み現象の定量評価

On the gas-liquid interface of a sodium-cooled fast reactor, gas (bubble) entrainment can be caused by free surface vortices, which may result in, e.g. disturbance in core power. Therefore, it is important to develop an evaluation model to predict accurately the entrained gas flow rate for the prevention of such undesirable phenomena. In the previous studies, simple gas entrainment experiments have been conducted, in which, typically, the gas entrainment is caused by free surface vortex in an upper-tank and the entrained gas is separated from liquid in a lower-tank which is connected with the upper-tank by a suction pipe. The entrained gas flow rate had been measured while the liquid flow parameters, e.g. liquid level or flow rate, are changed. However, the influence of the pressure difference between the upper and lower tanks had not been focused on those studies. In this study, to measure accurate the entrained gas flow rate, a simple gas entrainment experiment is conducted with focusing on the effect of the pressure difference between upper and lower tanks. The pressure difference between upper and lower tanks are controlled by changing the gas pressure in the lower-tank. As a result, it is confirmed that the entrained gas flow rate increases with increasing the pressure difference between upper and lower tanks. By the visualization of the two-phase flow in the suction pipe, it is also observed that the pressure drop in the suction pipe increases with the increase in the entrained gas flow rate when intermittent liquid plug is generated in the two-phase flow.

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Abstract Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the nuclear power plant (NPP) Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 libraries are considered. In the first part of this work the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few 100 pcm in multiplication factor for a fresh fuel due to differences in 235U and 238U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout, and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1%. Some differences in isothermal temperature coefficient were observed, however, they only marginally affect core power defect going from zero to full power.

Neutronic studies for the second Egyptian research reactor ET-RR-2

Abstract The start up physics studies SPS of the Second Egyptian Research Reactor is presented in this work. The SPS consists of the approach to criticality, the power and flux distribution of the first critical core, criticality, power and flux distribution of the working core. The working core is the beryllium-reflected core with cobalt loaded for irradiation. Some of the neutronic parameters are determined, these parameters are the power peaking factor, power distribution, the excess reactivity for both the water and beryllium reflector cores, the core excess reactivity with the cobalt loaded, the reactivity worth of the cobalt irradiation box. Flux distribution for one fast, three eipthermal and one thermal energy group are also determined. All the above neutronic parameters are calculated by using the lattice cell code WIMSD4 and the two dimensional diffusion code DIXY2.

Modelling research on the control scheme and control characteristic of a small gas-cooled reactor

The nuclear power system is one vital choice for interplanetary migration. The control rod and sliding reflector are the two most widely used control schemes with different characteristics for the space nuclear reactor. In the present paper, the control rod scheme and sliding reflector scheme of the 2.6 MWt reactor model are investigated and compared. According to the neutron physics analysis, the core arrangement and the control assembly motion scheme for the reactor start-up process are proposed, along with the reactivity insertion behavior. The results show that the reactivity insertion of the control rod scheme is $18.76, which is lower than the $26.01 of the sliding reflector scheme, and the weight of the former is 79.6 kg larger than the latter. However, the control rod scheme has significant advantages in core power flattening and control accuracy. Upon comprehensive consideration, the control rod scheme is recommended as the control scheme. The research results can lay the foundation for the control system layout scheme and control strategy design.

Validation of the thermal-hydraulic thunder code and application for PWR core in transient conditions

The THUNDER computer code allows performing thermal-hydraulic core analysis of pressurized water reactors aiming at the design and safety analysis of power and naval reactors in steady state and transient conditions. It uses the subchannel and control volume methods to perform a tridimensional analysis of the nuclear reactor core. The conservation equations are solved simultaneously for each control volume based on geometric and operation conditions. The Integral Momentum Method is used for solving the conservation equations decoupling the momentum equation from the mass and energy conservation equations. The drift-flux model is adopted to treat the two-phase flow condition. The analysis is performed in two-stages: first, a macroscopic analysis is conducted with subchannel equivalent to a fuel assembly (bundle), and second, with a subchannel equivalent to an individual coolant channel. Only the hottest fuel assembly is analyzed with detail. To validate the THUNDER code we used the experimental data from the NUPEC PWR benchmark (subchannel and bundle tests), in which we verify the accuracy of the coolant temperature at steady-state in 4 different void fraction conditions and the accuracy of the void fraction at transients initiated by four different conditions: core power increase, reduction of inlet mass flow rate, increase of inlet coolant temperature and depressurization of the inlet coolant. At steady-state conditions, the mean relative discrepancy for the coolant temperature was 0.09. At transient conditions, the void fraction mean absolute discrepancy for power increase was 0.045, for mass flow reduction was 0.079, for inlet coolant depressurization was 0.032 and for temperature increase was 0.092. The code was applied for the Brazilian 1350 MW nuclear power plant ANGRA 2 for evaluating the thermal-hydraulic safety parameters at steady-state, as well as the operational transients of loss of one or several primary coolant pumps. The THUNDER code results showed that the safety limits were not exceeded in concordance to the Final Safety Analysis Report of the plant. These results indicate that the THUNDER code is sufficiently accurate to perform thermal-hydraulic analyses for PWR power plants at steady-state and transient conditions.

Numerical investigation on flow-induced vibration and heat transfer of fuel rods in a small pitch-diameter ratio (P/D) channel

The Korea Atomic Energy Research Institute developed a dual-cooled annular fuel rod bundle structure to enhance the core power of the nuclear reactor and improve its heat transfer efficiency. The structure also reduced the pitch-diameter ratio (P/D) of the channel. In this study, a four-subchannel model is taken as the research object, and the Reynolds stress model (RSM) is applied to calculate the flow-solid-thermal tri-physical field coupling for a typical bare rod beam channel. The accuracy of the numerical calculation method used in this study was first verified by comparing it with the experimental data in the literature. Then, the channel flow field under different P/D conditions was simulated, and the formation conditions and influencing factors of the vortex-street structure were studied. Axial flow-induced vibrations of a cylindrical structure in a channel were investigated using a weak coupling algorithm. Finally, the influence of the vortex-street structure on the heat transfer efficiency of the fluid in the channel is discussed. The results show that the anisotropy-based RSM turbulence model can accurately capture the vortex-street structure in small P/D channels. When P/D < 1.050, a stable vortex-street structure appeared in the channel, which led to the stable micro-amplitude vibration of the cylindrical structure in the channel. The presence of the vortex-street structure significantly improves the heat transfer efficiency of the fluid in the channel.

Study on Prognostics and Health Management of Fluid Loop System for Space Application

Abstract Prognostics and Health Management (PHM) is an important technical means to realize condition-based maintenance of complex systems. As a complex space application engineering, the space application fluid loop system has the requirements of high safety, high reliability and long life. Therefore, this paper adopts PHM technology to comprehensively study the space application fluid loop system from two aspects of architecture and key technology respectively. Firstly, this paper introduces the current status of health management of fluid loop system for space application and the problems to be solved urgently. Secondly, the PHM system architecture and key technologies such as condition monitoring, health assessment, fault diagnosis, fault prediction and health management are described, and the PHM system research scheme of fluid loop for space application is established. Finally, a case study is carried out by taking the fluid loop main engine of the core power equipment as an example. The analysis results show that the PHM technology can effectively reduce the system failure rate, improve the system safety and reliability, and provide a strong guarantee for the long-term on-orbit operation of the system.

A preliminary analysis on optimization of core flow distribution

Optimization of core flow distribution is closely related with thermal safety, and a good match between core flow distribution and power distribution in reactor core is helpful for the promotion of thermal-hydraulic design, especially for the reactor with natural circulation whose driving force is limited. Flow distribution can be contrived by the reasonable resistance arrangement of core inlet in reactor design. In this study, an automatic solution search algorithm on optimizing inlet resistance distribution of reactor core is designed and established, and also an automatic core flow distribution optimization code coupled with COBRAIIIC code is developed in place of the method by skill and experience of user. A natural circulation reactor is selected to be tested, and the results show that the coolant temperature difference of assembly channel outlets at a specific time point of fuel cycle life can be eliminated by optimizing the core flow distribution. Further, for the flow distribution optimization during fuel cycle life which could be regarded as a global optimization problem with range limits of independent variable, simulated annealing algorithm is applied to solve it. A good optimization result that the coolant temperature difference of the assembly flow channel outlets during a fuel cycle life is reduced effectively is obtained, which provides a support for further enhancement of reactor system performance.

High-Fidelity Steady-State and Transient Simulations of an MTR Research Reactor Using Serpent2/Subchanflow

In order to join efforts to develop high-fidelity multi-physics tools for research reactor analysis, the KIT is conducting studies to modify the coupled multi-physics codes developed for power reactors. The coupled system uses the Monte Carlo Serpent 2 code for neutron analysis and the Subchanflow code for thermo-hydraulic analysis. Serpent treats temperature dependence using the target motion sampling method and Subchanflow was previously extended and validated with experimental data for plate-type reactor analysis. This work present for the first time the steady-state and transient neutron and thermo-hydraulic analysis of an MTR core defined in the IAEA 10 MW benchmark using Serpent2/Subchanflow. Important global and local parameters for nominal steady-state conditions were obtained, e.g., the lowest and highest core plate/channel power/temperature, the radial and axial core power profile at the plate level, and the core coolant temperature distribution at the subchannel level. The capabilities of Serpent2/Subchanflow to perform transient analysis with on-the-fly motion of the control plates were tested, namely with fast and slow reactivity insertion. Based on the unique results obtained for the first time at the subchannel and plate level, it can be stated that the coupled Serpent2/Subchanflow code is a very promising tool for research reactor safety-related investigations.

Vibration analysis for predictive maintenance and improved reliability of rotating machines in ETRR-2 research reactor

Abstract In this work, both hardware and software modifications in a typical research reactor protection system (RPS) is proposed. The reactor cooling pumps are tripped based on vibrations safety signals of the pumps while the reactor SCRAM signal is initiated based on low flow rate and pressure drop across the reactor core which is a direct result of pumps trip. The main objective of this work is to develop reactor SCRAM signal based of core cooling pumps vibration signals. The early shutdown of the reactor based on pumps vibration signals is of significant importance not only in cooling the decay power of the reactor core after shutdown but also to prevent pumps failure. In the hardware model, the core cooling pumps vibration signals are feed to RPS to initiate reactor SCRAM signal. In the software model, a modular artificial neural network (ANN) is used in modeling the vibration monitoring of the research reactor (ETRR-2). The input and the output signals of the vibration transducer are used as a source data for training the neural network model. The type of the network used in this methodology is the supervised Multilayer Feed-Forward Neural Networks with the back-propagation (BP) algorithm. Vibration analysis programs are used in research reactors (RRs) to identify faults in machinery, plan machinery repairs, and keep machinery functioning for as long as possible without failure. The vibration severity limits are determined based on the International Organization for Standardization (ISO) 10816. The ANNs were designed using two different methods; one is by using hardware application contained two out of three voting and dynamic modules for trip signal by using ANNs. The current model classifies the vibration signals into five ranges low, good, satisfactory, unsatisfactory, and unacceptable vibration. The ANN is trained to detect the signal and vote to take the correct and safe action. The results demonstrate that the ANN can help in taking predictive actions for the safe core coolant pumps operation.

Model predictive and fuzzy logic controllers for reactor power control at Reaktor TRIGA PUSPATI

The Reaktor TRIGA PUSPATI (RTP) rely extensively on a core power control system for control reactivity to fulfilment the fundamental safety function for a nuclear reactor. It is technically challenging to operate within tight multiple parameter constraints and keep the core power output stable. At present, the power tracking performance of the system could be considered unsatisfactory which produces a relatively long settling time and high control effort. Hence, a study of a model predictive control (MPC) strategy by integrating the current Power Change Rate Constraint (PCRC) using fuzzy logic which is part of the core power control design is conducted. In this paper, the MPC design based on mathematical models of the reactor core included point kinetics model, thermal-hydraulic model, reactivity model and dynamic rod position model help to enhance core power control. The power tracking performance of the proposed control method and previous Feedback Control Algorithm-Fuzzy PCRC is compared via computer simulation with different power range operations. Overall, the results show the MPC-Fuzzy PCRC strategy provides a greater level of operational safety and optimum operation of the TRIGA reactor.

Facies-Controlled Modeling for Permeability of Tight Gas Reservoir Based on Hydrodynamic and Geophysics Characteristics

Due to poor physical properties and strong heterogeneity of Daniudi tight gasfield (China), traditional methods are not ideal for predicting reservoir permeability. Based on geoscience data mining algorithms and modeling techniques, this parameter is predicted and characterized from a new perspective. The high precision hydraulic unit information was outputted by the BP neural network and the interwell database was predicted by sequential indicator simulation based on the logging and core data, the exponential and power law relationship functions were selected to participate in the construction of the permeability model. The results showed that the facies-controlled modeling system for permeability based on the hydrodynamic characteristics of tight gas reservoirs could effectively improve the accuracy of reservoir prediction, and the logging information in the longitudinal direction and the facies information in the plane were combined by the hydraulic unit.

Author’s Response

Abstract The author thanks the panelists for their comments. He appreciated their raising questions in need of further study and he tried to answer them: the concept of core power, security fringe, limits of Russian expansion, and he himself raised additional questions. One of the most intractable questions raised by Barbara Skinner concerned the role of the Jesuits in the conflict between the Orthodox and Catholic churches; another concerned the meaning of the phrase “the partitions of Poland.” The author answered them hopefully to her satisfaction. Semyonov placed the rise of the Russian Empire in the broad context of Eurasian geopolitics and the author responded by submitting a number of issues in need of further study. He only regrets that no one raised the broad issue of the roleof the law and its codification as a factor of integration, and expresses the hope that this panel will raise an awareness among interested scholars of strategic and economic issues.

A Study of Gamma Heating, Neutron Flux, and Gamma Flux Under Several Reactor Core Conditions in the McMaster Nuclear Reactor

Changes in the thermal power of a nuclear research reactor will lead to changes in experimental, irradiation, and testing conditions. Consequently, reactor core parameters are inevitably susceptible to changes. One such parameter is gamma heating (GH), which results from gamma interaction with materials. In this work, a gamma thermometer was used to measure GH over the course of 7 operational days and nights. In addition, the Monte Carlo reactor physics code Serpent-2 was used to evaluate the sensitivity of common detection methods for monitoring reactor core parameters such as neutron fluxes, GH, and gamma flux under the following conditions: reactor core power variation, reactor core fuel shuffling, and detector vicinity fuel assembly shuffling. The GH values obtained through measurements and calculations were linearly proportional to the reactor power. In addition, the Serpent-2 code for the McMaster nuclear reactor showed that despite maintaining the reactor power core at the same level, the fuel burnup distribution could alter the studied parameters.

Uncertainty quantification of the power control system of a small PWR with coolant temperature perturbation

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

Non-integer Order Control Scheme for Pressurized Water Reactor Core Power

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

Spirit (精神 jingshen) as a key contemporary Chinese concept

The ancient concept of spirit 精神 (jingshen) occupies a pivotal place within contemporary Chinese political discourse for which there is no equivalent in the West. Older Christian and subsequent Enlightenment usage of spirit, culminating in Hegel’s thought, was displaced by crude materialism, utilitarianism and scientific positivism. The Chinese spirit today is based in the thought of Mao, Deng and Xi, inspires patriotism and combines the socialist core values, path, power and discourse. All kinds of human action exhibit spirit. It is manifested in the Party, in the Long March, by Comrade Jiao Yulu, by Shanghai city, by space exploration. In public discourse, three main types may be distinguished, the exemplary individual, the paradigmatic deed and the collective inspiration. Spirit in China provides the connectedness of community and common purpose that the West struggles to rediscover. In responding to global challenges together, China and the West may share a spirit that transcends their differences.

Pin by pin simulation of VVER core power distribution under transient conditions

Pin by pin simulation of VVER core power distribution under transient conditions

Uncertainty Propagation Analysis of Ionization Chamber Sensitivity in Core Power Control System of Small Pwr

Uncertainty Propagation Analysis of Ionization Chamber Sensitivity in Core Power Control System of Small Pwr