Coolant Loop Research Articles

Optimizing coolant loop design across the stator core: A research study

Liquid cooling is widely used on high power motors. Optimal design of the coolant loop could help rise the power density. The coolant channels are placed across the stator core in this research. The heat dissipation performances of the coolant channel with different designs are compared and analyzed through simulation. Experiments were carried out to verify the simulation results and the feasibility of the cooling method. The results show that the coolant loop should be placed tightly close to the slots. The heat dissipation performance of the optimized coolant loop is efficient compared to the jacket cooling design. The coolant channels across the stator core could remain sealed under a 0.5 Mpa pressure at least according to the air proof test. The coolant pressure within the channels could be reduced efficiently through increasing the parallel loop number or moving the channels away from the slots properly. The winding temperature of the measured values and the simulated results of the motor with jacket cooling is within 2 °C.

Preliminary tritium transport analysis for the water cooled ceramic breeder blanket system based on a system level code

This paper presents a dynamic analysis tool for evaluating the tritium inventory, tritium extraction and tritium losses for the Water Cooled Ceramic Breeder (WCCB) blanket system of Chinese Fusion Engineering Test Reactor (CFETR). We have developed a program for the tritium transport analysis at the system level, which aims to estimate the tritium related quantities for the WCCB blanket system. The tritium inventory in the material defects is likewise taken into account. The sensitivity analysis is conducted on key parameters is carried out. The key output quantities of the WCCB blanket system include the tritium inventory in structural material, purge gas and coolant loops, tritium extracted by the TES and CPS, and tritium losses.

Verification of the SAM Code by Means of the Method of Manufactured Solutions: Application to Molten Salt Reactor Multiphysics Simulation

Software verification and validation constitute crucial phases in the development of simulation computer codes, particularly in the context of nuclear reactor safety analysis codes, where stringent safety requirements govern the development and deployment of nuclear technologies. This paper focuses on numerical verification study of the System Analysis Module (SAM) computer code, currently under development at Argonne National Laboratory. Specifically, we employed the Method of Manufactured Solutions (MMS) and proposed a verification technique tailored to the multiphysics simulation of molten salt reactors (MSRs). This research accomplished three main objectives. First, we have addressed key challenges associated with applying the MMS to MSR systems, arising from (1) the complex multiphysics coupling inherent in this problem and (2) the necessity to model the entire coolant loop for describing the drift of delayed neutron precursors outside the reactor core. The paper provides recommendations and guidelines to overcome these challenges, enabling the successful application of the MMS for simulating MSRs. Second, we have presented a comprehensive set of verification examples, serving as an exhaustive benchmark for code verification within the nuclear community. Third, we have established a robust verification of the SAM code’s capability to model MSR systems.

Fuel cell temperature control based on nonlinear transformation mitigating system nonlinearity

The operational temperature significantly impacts the efficiency and lifespan of fuel cells. However, precise fuel cell temperature control poses a challenge due to the inherent nonlinear characteristics within the thermal management system. In this study, an in-depth quantitative analysis is undertaken to dissect the intricacies of nonlinearity embedded in temperature control. The results show that the nonlinearity arises from the nonlinear mapping between the thermostat opening and the distribution of coolant flow between the inner and outer loops. To address this nonlinearity, a pre-experiment is implemented to determine the flow ratio of the outer loop coolant flow rate concerning the total stack coolant flow rate under different thermostat openings. Furthermore, a robust temperature control method based on nonlinear transformation is designed, ensuring that stack coolant inlet temperature quickly tracks the target value when load currents change. The proposed control method is verified on a 5 kW fuel cell test bench, and the results demonstrate improvement in dynamic characteristics and enhanced system's robustness. Small-step tests indicate stable tracking of the stack coolant inlet temperature, with a steady-state error of less than 0.2 °C. Besides, under a large-load step response test, the thermostat exhibits a rapid response with merely a 0.6 °C temperature overshoot.

A comprehensive analysis of thermal–hydraulic signatures in neutron noise of WWER-type reactors

The neutron noise phenomenon, occurring in all types of nuclear reactors, provides valuable insights into the dynamic behavior of these reactors. In this work, a well-justified approach for modeling neutron noise induced by thermal–hydraulic sources in a hexagonal reactor core is presented in detail. These sources encompass fluctuations in key features of the inlet coolant, including flow rate, temperature, and boric acid concentration. Our proposed approach touches on various aspects of the distribution of thermal–hydraulic parameters among coolant loops, accounting for both uniform and non-uniform distributions. The latter allows for the asymmetric and asynchronous distribution of their fluctuating component among the coolant loops. To manage such complexity, it becomes imperative to determine the contribution of each fuel assembly based on each coolant loop. To achieve this, CFD simulations are conducted for the downcomer and lower plenum of the reactor pressure vessel. These simulations provide essential data to establish the coolant mixing matrix, which accounts for non-uniform distributions. The obtained mixing matrix is subsequently integrated into our neutron noise simulator, DYNOSIM, ensuring the faithful reproduction of neutron noise. Our study comprehensively analyzes both uniform and non-uniform scenarios. The radial and axial distributions of neutron noise amplitude and phase are examined for these scenarios. Their qualitative verifications are conducted by prior works in this field. Furthermore, we compare one of the uniform scenarios with reference results obtained using a frequency domain simulator. Notably, our CFD simulation results closely match the experimental data. Our findings demonstrate the effectiveness and efficiency of this methodology in modeling and understanding neutron noise in hexagonal reactor cores.

Thermoelectrics for nuclear fusion reactors: opportunities and challenges

In this review, we discuss the promising applications and practical considerations of thermoelectrics to harvest the unutilized thermal gradient between the plasma-facing surfaces and the molten salt coolant loop in tokamak fusion reactors.

Safety Analysis in AP-1000 Nuclear Power Plant Against Primary Loop Coolant Pump and Fuel Failures Authors

Safety Analysis in AP-1000 Nuclear Power Plant Against Primary Loop Coolant Pump and Fuel Failures Authors

Analysis of Rotor-Seizure-Induced Pressure Rise in a Nuclear Reactor Primary Cooling Loop

Analysis of Rotor-Seizure-Induced Pressure Rise in a Nuclear Reactor Primary Cooling Loop

Assessment of Low Global Warming Potential Refrigerants for Waste Heat Recovery in Data Center with On-Chip Two-Phase Cooling Loop

Assessment of Low Global Warming Potential Refrigerants for Waste Heat Recovery in Data Center with On-Chip Two-Phase Cooling Loop

Corrosion protection and failure mechanism of ZrO2 coating on zirconium alloy Zry-4 under varied LiOH concentrations in lithiated water at 360 °C/18.5 MPa

Maintaining a good corrosion resistance in primary coolant loop during normal operations is a prerequisite for ZrO2 as a protective coating on zirconium alloy cladding. Research on corrosion performance of ZrO2 coating in nuclear water chemistry is relatively deficient, and existing reports failed to provide an in-depth explanation for the failure causes of ZrO2 coating. Here, we proposed a detailed corrosion process of ZrO2 coating in lithiated water at 360 °C and 18.5 MPa based on experimental study and molecular dynamics simulation. The protective effect and failure mechanism of ZrO2 coating on Zry-4 under varied LiOH concentrations was further revealed. ZrO2 coating provided a favorable corrosion protection with the occurrence of localized corrosion at low LiOH concentrations. Factors influencing the corrosion resistance include pitting corrosion extension, enhanced Li+ permeation, short-circuit diffusion of O2– and ZrO2 phase transformation. In highly-concentrated LiOH solutions, intergranular corrosion, internal oxidation and perforation result in coating failure. Zr ions were released to the coating surface to form flocculent ZrO2 and ZrO2 clusters due to the strong diffusion and dissolution tendency of α-Zr in the Zry-4 substrate. This work can give some references to understand the service behavior of nuclear coatings under variable water chemistry conditions and promoting in-pile application of ZrO2 coating.

Method for predicting lifetime of ion exchange resin in PWR primary loop based on nuclide distribution measurement

Ion exchange resins are commonly utilized for water treatment in pressurized water reactors' primary loop. These resins have the ability to adsorb radioactive cations present in the primary loop coolant, leading to a purification effect. However, the resin becomes radioactive post-purification and cannot be regenerated, thus requiring regular replacement. As the process of resin replacement exposes personnel to radiation, it is crucial to enhance the resin's utilization. This paper aims to investigate a non-destructive testing method for evaluating the lifetime of resin in pressurized water reactor purification units. It also aims to offer guidance for safe resin replacement. The resin is contained in a column (commonly referred to as ion exchange columns) and consumed from top to bottom, with radioactivity decreasing in intensity from top to bottom. The study utilizes the longitudinal radioactivity distribution characteristics of ionic resins to build a gamma-ray detection system with high resistance to interlayer interference and good longitudinal resolution. In this study, a two-stage resin column was utilized to capture common fission product nuclides and corrosion-activated nuclides, such as 60Co, 137Cs, 54Mn, and 124Sb, from the effluent of a pressurized water reactor. The distribution of these nuclides was measured using this device. The results indicate that the activity levels of 60Co, 137Cs, and 54Mn reached their peak in the column before rapidly declining. The height of the resin layer corresponding to the highest activity level of each nuclide varied, indicating the resin's adsorption selectivity. After processing an additional 10 tons of wastewater, a second measurement was taken which showed that the failure range of the resin expanded downwards as the treated water volume increased. The entire distribution curve of the nuclides within the exchange column also shifted upwards and to the right. By comparing the first and second observations, we discovered that the bottom part of the resin will fail when the activity of the most penetrating nuclide in the wastewater surpasses a specific threshold. This research suggests a novel approach to predicting the lifetime of resin, which generalizes the trend of radionuclide intensity inside the ion exchange column.

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. To achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.This manuscript has been authored in part by UT-Battelle, LLC, under contract DE AC05-00OR22725 with the US Department of Energy (DOE). The publisher acknowledges the US government license to provide public access under the DOE Public Access Plan (https://energy.gov/downloads/doe-public-access-plan [energy.gov]).

On the applicability of advanced model-based strategies to control of electrified vehicle thermal systems

Electrified Vehicles (xEVs) often rely on complex thermal systems to meet energy efficiency and performance targets. These systems are typically made up of multiple interacting loops, such as coolant loop, oil loop, and refrigerant loop which can be highly nonlinear, strongly coupled, and teemed with multiple sources of uncertainties and disturbances. Designing effective control strategies for such complex thermal systems is a difficult and crucial task. This paper proposes three advanced Model-Based Control (MBC) strategies for the cabin active heating thermal system of an EV, namely, a Nonlinear Model Predictive Control (NMPC), a Model Predictive Control via Value Function Approximation (MPC-VFA), and a Linear-Quadratic Regulator (LQR). For this purpose, a physics-based model of the aforementioned thermal system is derived and validated using data generated from a high-fidelity thermal plant model created in GT-suite software. The parameters of the physics-based model are identified using the particle swarm optimization algorithm and it is shown that the nonlinear dynamics of the thermal system has been accurately captured. A comprehensive test study is performed and the performance of the proposed MBC approaches is evaluated using various indicators, such as reference tracking, energy consumption, robustness, computation time, and implementation complexity. Experimental data has been also utilized to validate the high-fidelity thermal plant model. The results of the test study demonstrate a high performance and efficiency of the proposed control strategies, offering significant advancements in thermal system control of Electrified Vehicles.

Transient thermal–hydraulic analysis of the thermionic space reactor TOPAZ-II

Space nuclear reactor system is of great significance in aerospace applications such as deep-space exploitation and exploration, high-power earth-orbiting satellites. A transient reactor system analysis code, RESYS, is developed for advanced non-light water reactor safety analysis recently. In this paper, the model of the thermionic space nuclear reactor TOPAZ-II including the thermionic fuel elements, thermionic energy conversion system, reactor core, coolant loop and radiator is established with RESYS code. The point reactor equation with six-group delayed neutron is utilized to calculate the fission power. The reactivity effect of various structural components and the reactivity insertion effects of control drums are considered. The thermionic emission model of cesium thermionic converter is based on the phenomenological model established by Rasor. Then, the steady state operation and three transient accidents including reactivity insertion accident (RIA), partial loss flow accident (LOFA), load resistance shortcircuit accident (LRSCA)are simulated. In the steady-state condition, the output electric power is about 5.8 kW with thermoelectric conversion efficiency of 5.0%. The results of steady calculation agree well with the design values. And in the three transient accidents, the moderator introduces a large temperature reactivity feedback which makes the reactor fission power increasing. The results indicate that the TOPAZ-II reactor system has a sufficient safety margin during these accidents. In addition, the calculation results preliminarily demonstrate the capability of RESYS code in modeling and simulation of thermionic space reactor.

Thermal Performance of the Evaporator in the two Phase Mechanically Pumped Cooling Loop

In the two-phase mechanically pumped cooling loop, the performance of the evaporator is very important as it directly contacts with the electronic equipment to absorb the waste heat. The evaporator with the micro channels which can enhance heat exchange is designed and it can be used to dissipate the heat generated by the electronic equipment with discrete heat sources. The effects of working medium flow rate and heating power on the performance of the evaporator were analyzed through numerical simulation. Meanwhile, the changes of evaporator performance were also investigated while the heating power changed periodically and suddenly. In addition, the flow and heat transfer process in the evaporator at the initial stage were analyzed by transient calculation. The results showed that the evaporator with micro channels can reduce the maximum temperature by 47.66% and the maximum temperature difference by 83.39% compared with the evaporator without micro channels. In addition, the evaporator showed good stability under variable power condition.

Loss Of Flow Accident (LOFA) with protection in NUR nuclear research reactor; three dimensional analysis of a Fast LOFA and a Slow LOFA

The Loss Of Flow Accident (LOFA), with protection and without intervention of natural convection in the cooling process, is studied in the NUR nuclear research reactor for two accident configurations; a Fast Loss Of Flow Accident (FLOFA) and a Slow Loss Of Flow Accident (SLOFA). The accident is due to an inadvertent partial closure of coolant loop isolation valve. All the sequences of each accident are presented, namely the steady-state, the transient (SLOFA or FLOFA), the nuclear reactor emergency shutdown (SCRAM), the cooling of the decay heat and the contribution of 20% of flow to guarantee a good nuclear reactor core cooling. In order to be able to follow the thermalhydraulic accident down to the smallest detail, the equation set-up, based on the three conservation equations of continuity, momentum and energy, in three dimensions of space and one dimension of time, coupled to the turbulence model K−ω SST, is established. The models of the distribution of thermal power and decay heat are introduced into the mathematical model as well as the exponential flow loss equation. The boundary conditions for each sequence of the accident are fixed. The resolution of all the equations was carried out using Computational Fluid Dynamics (CFD) through FLUENT calculation code. It was possible to show the evolution of the temperature of the coolant and the cladding in the respective planes of each physical quantity for each time step of the accident, one second for the FLOFA and 2 s for the SLOFA. The evolution of the reactor power is presented before, during and after the SCRAM for each accident. After the SCRAM, the decay heat is taken into account and the average evolution of the temperature of the coolant and the cladding is presented as a function of time. The cooling of decay heat with no adding flow and with adding 20% flow was also studied as a function of time and the difference between the two approaches is presented. The FLOFA was studied for 2.2 s and the SLOFA for 25 s. The results obtained for the FLOFA and the SLOFA show that these accidents generates high temperatures but always remaining below the critical temperatures whether for the heat transfer fluid or for the cladding for the conditions and the data which have been retained for this work.

Score-based Bayesian network structure learning algorithms for modeling radioisotope levels in nuclear power plant reactors

Radioactive corrosion products released into the primary coolant loop dominate the final shutdown radiation fields of pressurized water reactors. Thus, reducing the concentration of these corrosion products is a paramount duty in the optimization process of the reactor performance. However, the complexity and uncertainty present in this process make it difficult to predict their evolution in a theoretical way. We propose the application of structural learning of Bayesian networks to discover the complex relations between the corrosion products and the most relevant variables in the primary loop, giving rise to probabilistic models that obtain accurate and reliable predictions of the corrosion products. Our analysis of 5 power plants demonstrates that our approach results in simpler and more reliable models. Additionally, we conclude that the learned structures may represent an interpretable tool for power plant technicians since they reveal useful information that can be directly employed to improve the reactor operation.

Experimental thermal-hydraulic characteristics of single-phase natural circulation loop using water-based hybrid nanofluids

Nanoparticle shape strongly affects the thermal-hydraulic behavior of nanofluid-enhanced single-phase natural circulation loop (SPNCL) but is yet to be explored experimentally. Hence, various mono/hybrid nanofluids having dissimilar shaped nanoparticles (Al2O3+Water, Al2O3+CuO + Water, Al2O3+SiC + Water, Al2O3+MWCNT + Water) with a 0.1% volume concentration, are used in SPNCL to investigate mass flow rate (MFR), heat exchanger effectiveness (HEE) and total entropy generation rate (TEGR). The effect of power input, coolant inlet temperature and loop inclination (both counter-clockwise and clockwise) on transient and steady-state performances are presented. The result reveals that mono/hybrid nanofluid shows enhanced HEE and lower TEGR as compared to base fluids, whereas the MFR may increase or decrease since the performance of SPNCL is significantly impacted by the nanoparticle's shape. Spherical-shaped nanoparticle-dispersed nanofluids show early flow initiation and higher MFR than cylindrical-shaped nanoparticle-dispersed nanofluids. MFR and TEGR rise with higher power input, whereas HEE declines for all working fluids. MFR and TEGR increase as the coolant inlet temperature rises, whereas HEE decreases for all working fluids. The loop inclination decreases the MFR, while it increases the HEE and TEGR. The Counter-clockwise inclination of the loop shows a higher MFR than the clockwise inclination at the same angle of inclination.

Battery Thermal Management System for Electric Vehicles

Electrical vehicles (EVs) as a result of their rapid evolution and growing popularity, zero-emission, and high tank-to-wheelefficiency. Though, some features, particularly those relating to battery performance, cost, lifetime, and protection, restrict the development of the electrical car. In order to operate at peak efficiency under various circumstances, battery management is therefore required. The BTMS is essential for controlling the thermal performance of the battery. The BTMS technologies include heating, air conditioning, liquid cooling, direct refrigerant cooling, phase change material (PCM) cooling, and thermoelectric cooling. Performance, weight, size, cost, dependability, safety, and energy consumption are trade-offs analyzed for these systems. According to the analysis the system is made up of two coolant loops, one refrigeration loop, and one cabin HVAC loop. The batteries, drivetrain, and cabin all contribute to the thermal burden. The model of these system is been built in the software MATLAB/SIMULINK. Based on the outcomes of the simulation, BTMS is crucial for regulating battery thermal behavior. Through the integration of the simulation model with battery thermal and ML models, next research might be more thorough and precise.

3D field data of flow structures in dead-ended, coolant loop reactor branch lines for CFD validation of thermal fatigue onset modeling

Thermal fatigue phenomena in normally stagnant branch lines connected to the primary loop of pressurized water reactors has been identified as a primary cause for at least ten pipe cracking events in the past decade. Current prediction and inspection of at-risk piping is based on semi-empirical models, derived on the basis of low-resolution experimental data. This data does not include sufficient information on contributing factors that lead to thermal boundary development and fluctuation in dead-ended branch lines. Without comprehensive, high-resolution data of the phenomena involved, CFD models attempting to account for the re-laminarization or quasi-laminar flow that persists in the branch line cannot be validated.Validated CFD models for dead-ended branch line flow conditions will greatly expand the predictive capabilities the method currently offers. To build the database necessary for related CFD validation campaigns, 3D high-resolution measurements were made in the branch line of an experimental facility built at the University of Michigan. The facility is a scaled coolant flow loop with a reduced main line diameter. Nuclear power plant coolant loop flow rates near 10 m/s were achieved for study. High-resolution, 3D velocity field data in the dead-ended branch line was obtained using Tomographic Particle Image Velocimetry (T-PIV) techniques. With this data as a basis for validation, CFD models such as Low-Re will be able to accurately predict flow phenomena in dead-ended branch line simulations where the branch line diameter, opening geometry, or piping component (elbow) locations are altered.