Reactor Coolant Temperature Research Articles

Comparison study of two control strategies for the small pressurized water reactor

The control strategy has a significant impact on the control performance of the reactor system. This paper investigates the feedwater flow and steam flow control strategies for small pressurized water reactors. First, the mathematical model of the SPWR Nuclear Steam Supply System is introduced. Second, a dual-constant control strategy was proposed, and the feedwater flow rate and steam pressure control systems under the two control modes were designed. Moreover, a steam bypass control system was designed to avoid excessive steam pressure under the feedwater flow control mode, and the reactor power and pressurize control systems were introduced. Simulation results indicate that the two control modes both meet the SPWR control requirements. However, the feedwater flow control mode has better reactor coolant temperature and steam pressure control performances during the load rejection transient, owing to the timely discharge of heat from the reactor coolant system through the steam bypass system.

Machine Learning Based Fault Classification in Pilot Plant Batch Reactor: Using Support Vector Machine.

Identifying and diagnosing faults is a critical task in process industries to maintain effective monitoring of process and plant safety. Minimizing process downtime is critical for enhancing the quality of the product and minimizing production costs. Real-time categorization of issues across several levels is essential for the monitoring of processes. However, there are still notable obstacles, that must be addressed, such as the existence of robust correlations, the complexity of the data, and the lack of linearity. This study introduces a novel fault identification technique in batch reactor experimental trials that employs multikernel support vector machines (SVMs) to categorize internal and external issues, specifically reactor temperature, coolant temperature, and jacket temperature. The data set was obtained from empirical research. The classification has been conducted using a multikernel SVM. This article identified that the nonlinear classifier using the radial bias function results in an accuracy that is at least 22.08% superior to other methods.

Conceptual design and analysis of a combined passive cooling system for both reactor and spent fuel of the pool-vessel reactor system

A novel combined long-term passive cooling system is proposed for the pool-type reactor, aiming to enhance the safety and reliability of the entire system during extended station blackout accidents (SBO). This passive cooling system primarily consists of a natural circulation primary loop, a C-type heat exchanger submerged in the pool, the reactor pool (RP) and the spent fuel pool (SFP), and a two-phase loop thermosyphon (TPLT) cooling subsystem. The TPLT cooling subsystem comprises a separated evaporator and condenser, with the evaporation bundle immersed in the SFP and the condensation bundle placed within a dry air cooling tower. Three typical design configurations were selected for numerical simulation and performance evaluation, including C-type tube bundles placed inside the RP and TPLT evaporation bundles placed inside the SFP, C-type tube bundles and TPLT evaporation bundles both placed inside the SFP, and C-type tube bundles and TPLT evaporation bundles both placed inside the RP. The analysis results demonstrate that this combined passive cooling system can effectively achieve passive cooling of the reactor and spent fuel under accident conditions, mitigate the rise in reactor coolant temperature, and prevent saturation boiling in the RP and SFP. Furthermore, different system layout configurations have a significant impact on its operational characteristics, allowing for rational optimization and improvement based on specific safety requirements of the system.

A Raman spectroscopic and ab initio investigation of aqueous boron speciation under alkaline hydrothermal conditions: evidence for the structure and thermodynamic stability of the diborate ion.

Raman spectra of aqueous sodium borate solutions, with and without excess NaOH, NaCl, and LiCl, have been obtained from perpendicular and parallel polarization measurements acquired using a custom-built sapphire flow cell over the temperature range 25 to 300 °C at 20 MPa. The solvent-corrected reduced isotropic spectra include a large well-defined band at 865 cm-1 which overlaps with the boric acid B(OH)3 band at 879 cm-1, and becomes increasingly intense at elevated temperatures. This band does not correspond to the spectrum of any other previously reported aqueous polyborate ions, all of which have symmetric stretching bands at frequencies below that of borate, [B(OH)4]-, at 745 cm-1. Based on the classic high-temperature potentiometric titration study by R. E. Mesmer, C. F. Baes and F. H. Sweeton, Acidity measurements at elevated temperatures. VI. Boric acid equilibriums, Inorg. Chem., 1972, 11, 537-543, the new band was postulated to arise from a diborate ion, [B2(OH)7]- or [B2O(OH)5]-. Ab initio density functional theory (DFT), together with chemical modelling studies, suggest that it is most likely [B2(OH)7]-. Thermodynamic formation quotients derived from the peak areas showed variations with ionic strength as well as charge-balance discrepancies, which suggest one or more unidentified minor equilibrium species may also be present. The most likely candidate is the divalent diborate species [B2O2(OH)4]2- which is also predicted to have a band near 865 cm-1 and is postulated to be present as a sodium ion pair. These are the first quantitative Raman spectra ever reported for borate-rich solutions under such conditions and provide the first spectroscopic evidence of a diborate species at PWR reactor coolant temperatures.

Power-pressure coordinated control of modular high temperature gas-cooled reactors

Modular high temperature gas cooled reactor (mHTGR) adopts helium as coolant and graphite as both the moderator and structural material, whose fuel element is built by embedding thousands of TRISO coated particles into a graphite matrix. Since helium is an ideal gas, the dynamics of reactor coolant temperature is tightly coupled with that of primary pressure. To further improve the operation performance, it is meaningful to develop the coordinated control of power-level and primary pressure for mHTGRs. In this paper, a passivity-based power-pressure coordinated control is proposed for mHTGRs, and the sufficient condition for globally asymptotic closed-loop stability is given. This newly-built power-pressure coordinated control law is applied to the mHTGR of six-modular high temperature gas-cooled reactor plant HTR-PM600, and simulation results in the case of power ramping decrease and increase as well as performance comparison not only verify the theoretical result about closed-loop stability but also illustrate the influence of control parameters as well as the stronger performance relative to the classical proportional integral/differential (PI/D) controller.

Modelling and simulation of the primary system for a small lead-cooled fast reactor with a ratio of core power to flow

To study the operation strategy of primary system for small lead-cooled fast reactor, operation strategies including a ratio of core power to flow and stable core power are designed. First, the nonlinear model of primary system is transformed into transfer function model. Second, to maintain the stability of reactor power and coolant temperature, the operation strategies of stable core power and ratio of core power to flow are established with the PID controller. Finally, simulation of flow disturbance at the secondary side of heat exchanger is carried out. The results show that the core outlet temperature is excessive with the operation strategy of stable core power due to constant core power and can be basically kept constant with a ratio of core power to the flow model.

Dynamic modeling and two-degree-of-freedom controller design for pressurized water reactor core

The task of this investigation is to model pressurized water reactor core and design a reactor power controller that has a strong robustness in various operating conditions. The comparisons of different nodal models of the reactor core, which includes point kinetics, heat transfer between fuel and coolant, and reactivity feedbacks, indicate that the 2C/1F finite-difference nodal model with four fuel nodes has sufficient transient response accuracy. Based on the transfer function model that is obtained by linearizing the nonlinear model, a two-degree-of-freedom controller is proposed to reduce the effect of power level variation on the system dynamics. The simulations demonstrate that the proposed two-degree-of-freedom controller can give satisfying performance for both reactor power and coolant temperature over a wide range of reactor operation.

Performance analysis of cogeneration energy conversion system design for RDE

An Experimental Power Reactor called RDE is a non-commercial power reactor with the thermal power of 10 MW. RDE is being designed based on the high-temperature gas-cooled reactor (HTGR). The purpose of the RDE designed is to supply thermal energy for experimental purposes and to generate electricity for supplying the electricity in Serpong research area. The energy conversion system of RDE is designed in cogeneration configuration with a Rankine cycle. The outlet reactor coolant temperature of RDE is 700 °C, and the steam temperature at the outlet of the steam generator is 530 °C. The performance analysis of the RDE energy conversion system is concentrated on two parameters are the thermal efficiency and the energy utilization factor (EUF). This research aims to calculate and analyse the optimal of the two parameters in the RDE energy conversion system. Calculation and analysis of the performance of the RDE energy conversion system are performed by simulation using computer code ChemCad. The simulation results show that the energy utilization factor of the cogeneration for the RDE energy conversion system can increase by more than 60%. It can be concluded that the energy conversion system in a cogeneration configuration can increase the EUF.

MNSR transient analysis using the RELAP5/Mod3.2 code

To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated. The model was verified by comparing the results with the measured and the previously calculated data. The comparisons consisted of comparing the MNSR parameters under normal constant power operation and reactivity insertion transients. Reactivity Insertion Accident (RIA) for three different initial reactivity values of 3.6, 6.0, and 6.53 mk have been simulated. The calculated peaks of the reactor power, fuel, clad and coolant temperatures in hot channel were calculated in this model. The reactor power peaks were: 103 kW at 240 s, 174 kW at 160 s and 195 kW at 140 s, respectively. The fuel temperature reached its maximum value of 116 °C at 240 s, 124 °C at 160 s and 126 °C at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.

LQG/LTR Controller Design Based on Improved SFACC for the PWR Reactor Power Control System

The task of this investigation is to design a controller that has a strong robustness in various operating conditions. A new structure of state feedback assisted classical control (SFACC) that uses a differential lag compensator in the inner classical control loop is proposed to improve the robustness of original SFACC. The linear quadratic Gaussian with loop transfer recovery (LQG/LTR) at the plant output is employed to design the robust controller in the outer control loop. A comparison of the performance and robustness between the gain-insensitive controller and an existing LQG/LTR controller is made by nonlinear simulations. The proposed gain-insensitive LQG/LTR controller can give satisfying performance for both reactor power and coolant temperature over a wide range of reactor operations.

Intensified isothermal reactor for methanol synthesis

The performance of different tubular reactor designs for methanol synthesis was investigated by using computational fluid dynamics (CFD) simulations and experiments. The validated model was used to investigate the effect of various reactor geometries, coolant temperature, feed temperature, inlet composition, and flow rate on the methanol yield and heat transfer performance. New designs using metal inserts were proposed with an aim to improve the methanol yield by simple retrofitting. It was observed that the metal inserts can be utilized to relocate the catalyst close to the cooling surface, and as a result, near isothermal operating condition was achieved in both tube-cooled and tubular reactors. The methanol yield in tube-cooled and tubular reactors with the proposed metal inserts was improved by an average of 36% and 27% respectively over the range of space velocity.

Coupling simulation of neutron kinetics core model with CFD of IPWR steam line break accident

Abstract In this paper, development of coupled codes using two-group neutron diffusion kinetics code and computational fluid dynamics (CFD) solver Fluent has been introduced. Way of coupling, time step control algorithm and spatial mesh overlays have been summarized in detail which are basic components and challenges of the coupling methodologies. The implement and verification of coupled code have been modeled on integral pressurized water reactor (IPWR) IP200 with hexagonal fuel assembly in the core and once-through steam generators. The steam line break core transient was analyzed in coupled code simulation of a core boundary conditions derived from system code simulation results. The results presented transient three-dimensional distribution of the key operation parameters such as reactor power and coolant temperature, also demonstrated the inherent safety features of IP200. The current work will bring about the ability to explore multi-scale and multi-dimensional safety transient evaluations and give more precise neutronics/thermal-hydraulics mapping.

Analysis of passive residual heat removal system in AP1000 nuclear power plant

The paper presents operating principles of passive safety systems used in modern nuclear power plants with AP1000 reactor. Paper describes in detail the passive residual heat removal system and the passive containment cooling system in above-mentioned power plant. Furthermore, the paper consists thermo-hydraulic analysis of a scenario, in which power plant loss off-site power, reactor is shut-down and the active elements (especially Diesel generators) failed. Considerations and calculations present the role of passive safety systems in nuclear power plant. In particular described phenomena concerns two above-mentioned systems, which are responsible for residual heat removal. Change of the few of the most important parameters, like reactor coolant temperature, medium flow or heat exchange in the refuelling water storage tank, were simulated, presented and described. Mathematical model also includes the role of passive systems and its impact on the main parameters in nuclear power plant with AP1000 reactor.

Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

Using the Sound of Nuclear Energy

The generation of sound by heat has been documented as an acoustical curiosity since 1568 when a Buddhist monk reported in his diary the loud tone generated by a ceremonial rice cooker. Over the last four decades, significant progress has been made in understanding thermoacoustic processes, enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-powered and can wirelessly transmit the temperature and reactor power level by generation of a pure tone that can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor's coolant temperature and that the amplitude (above an onset threshold) is related to the reactor's operating power level. These signals can be detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

AGENT-TH scheme: New thermal hydraulics (TH) code coupled with the AGENT neutronics code for light water reactors modeling and analysis

In this paper a newly developed thermal hydraulics (TH) code coupled with the state-of-the-art neutronics deterministic code AGENT (Arbitrary Geometry Neutron Transport) is presented. This coupled system is called the AGENT-TH scheme. The neutronics AGENT code uniquely combines the method of characteristics (MOC) and R-function geometrical modeling to solve the 2D/3D neutron transport in nuclear reactors of current or future designs. Commonly, the cross sections are prepared using the SCALE code system. The implementation of the TH code with AGENT enables coupled thermal-hydraulic and neutronics analysis in providing the average reactor fuel temperatures. This allows for an accurate overall power profiling within the reactor system, which affects reactor coolant temperature, neutron flux, and other associated reactor parameters.The two main types of LWRs are boiling water reactors (BWRs) and pressurized water reactors (PWRs). The BWRs operate at lower pressures than PWRs, which results in a higher void fraction, or fraction of coolant flow which is vapor versus liquid. Performances of PWRs, with a higher operating pressure, may be obtained using the homogeneous equilibrium mixture (HEM) model, which assumes a slip relation equal to one between the coolant liquid and vapor phases. To account for the higher void fraction in the BWR coolant, the BWRs properties are assessed based on the drift flux model. The thermal hydraulics drift flux model is developed as a part of the TH code, to address the two-phase flow associated with two-phase heat transfer (including subcooled nucleation and saturated boiling) as found in BWR systems. This is accomplished based on the drift flux equations in 1D steady-state conditions, which permits fast computation times without sacrificing final accuracy.The TH code is benchmarked against the reactor simulation code TRACE for a single pin to assess the thermal-hydraulics capabilities of the drift flux model for the fluid void fraction, flow quality, pressure, saturation temperature, and mixture temperature along the axial height of the pin. The drift flux thermal hydraulics module is coupled with AGENT using a radial heat transfer model taking into account heat transfer from the fuel centerline to the bulk fluid. Fuel temperatures calculated in the TH code are used to update cross sections from the SCALE code system to account for thermal feedback. Reasonable agreements are obtained between the AGENT-TH and TRACE for the presented benchmark examples.

Modeling of SPERT IV Reactivity Initiated Transient Tests in EUREKA-2/RR Code

EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the experimental results provided by IAEA (International Atomic Energy Agency) obtained for a series of transient tests initiated by step insertion of different magnitudes of positive reactivity with varying degrees of different controlled parameters such as reactor initial power, coolant temperature and coolant flow condition. 20 out of 39 tests that fall under forced convection mode have been considered for the present simulation provided the reactor scram system is disabled. Peak power and peak clad temperature due to transient have been calculated and it was found that although peak clad temperature values agreed, the peak power values seem to underestimate the experimental values. Further study appears to be needed to identify the limitations in modeling or examining the effect of input parameters during modeling to obtain the better simulation results.

Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50–70 °C to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5 × 1025 m−2 to reach the total ion fluence of 1 × 1026 m−2 in order to investigate the near-surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate the irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near-surface (<5 µm depth) deuterium concentration increased from 0.5 at% D/W in 0.025 dpa samples to 0.8 at% D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near-surface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.3 dpa) at 500 °C cases even in the relatively low ion fluence of 1026 m−2.

Power-Level Control for MHTGRs with Time-Delay in Helium Temperature Measurement

The modular high temperature gas-cooled reactor (MHTGR), which has the inherent safety feature, high thermal efficiency and satisfactory economic feasibility, can be applied for electricity and process heat production. Power-level control is an important technique for providing both the stable operation and load-following performance. Since the coolant temperature sensors of an MHTGR are usually installed near the primary side of the corresponding steam generator, there must be time-delay effect in the feedback loop of the coolant temperatures. Moreover, the measurement signal transducing may also induce time-delay effect. Therefore, it is meaningful to give the power-level control design method by considering this time-delay effect. In this paper, a simple output-feedback power-level control is proposed for the MHTGRs by using the delayed measurement signal of average reactor coolant temperature. In the aspect of theoretical analysis, a sufficient condition, under which it is well guaranteed that this newly-built power-level control is a globally asymptotic stabilizer, is firstly given. In the aspect of verification, numerical simulation results not only verify the feasibility of the theoretical results but also show the relationship between the performance and values of parameters of this novel power-level controller. The meaning of this work lies in two aspects. The first one is deeply revealing the relationship between the closed-loop stability and values of the controller parameters. The second one is giving the approach of designing a simple and effective power-level control strategy to suppress the negative influence induced by the time-delay in the feedback loop of the coolant temperatures.

Theoretical and experimental modelling of a heat pipe heat exchanger for high temperature nuclear reactor technology

High temperature heat sources are becoming an ever-increasing imperative in the processing industries for the production of various plastics, fertilisers, coal-to-liquid fuel and hydrogen generation. Current high temperature reactor technology is capable of producing reactor coolant temperatures in excess of 950 °C. At these temperatures, tritium which is a radioactive contaminant found in the reactor coolant stream, is able to contaminate the secondary stream by diffusing through the steel retaining wall of the heat exchanger between the reactor coolant and secondary process coolant stream. Current regulations therefore require an extra intermediate heat transfer loop to ensure no cross contamination. A novel heat pipe heat exchanger design is presented which circumvents the need for an intermediate coolant loop. This is done by physically separating the reactor coolant and secondary coolant by two pipe walls and a vapour section and a liquid section. A theoretical transient heat transfer model of such a device is presented. The model uses separate hot gas heating fluid and cold water cooling fluid control volumes, and for the internal working fluid a control volume consisting of a liquid and its vapour in equilibrium with each other. A 2 kW rated experimental model was constructed and tested, using Dowtherm-A as working fluid, to validate the heat pipe heat exchanger theoretical model and design. By determining the boiling heat transfer coefficient through the use of an experimentally formulated correlation it was shown that the theoretical model is indeed able to simulate the characteristic chaotic behaviour, due to the boiling and condensation processes, of the device to within a reasonable level of accuracy. It is concluded that the theoretical simulation model can be used to predict the performance of a higher temperature sodium-charged heat pipe heat exchanger, provided suitable boiling and condensation heat transfer coefficients are used.