Coolant Loop Research Articles

Design modification of the existing cooling water system of Marsyangdi hydropower station using pipe flow modeling software

Marsyangdi Hydropower Station has adopted the Open Loop Cooling System (also called Once through Cooling System) in which the cooling water extracted from Penstock or the draft tube is circulated throughout the cooling system and discharged back to the draft tube. The water of the Marsyangdi River is muddy and sediment-rich in the rainy season, which causes the choking of tubes of the primary heat exchangers reducing the cooling efficiency and even leading to leakages in the tubes due to erosion. Cleaning and maintenance of the tubes are impossible without dismantling the primary heat exchangers from the Unit assembly. Hence, hours of machine shutdown is required to replace them with new or cleaned one. A Close Loop Cooling System has been modeled in the Software, "AFT Fathom," to overcome the issue in the Existing Cooling System. The modified model was run to observe the performance of the modified system at different river water temperatures and increased heat duty in all the heat exchangers. Further, the System Curve was considering the effect of the increase in surface roughness of pipelines due to aging. The performance capacity of the modified system was justified considering the obtained results from the simulations.

Vibration cause analysis and elimination of stagnant branch pipeline

Fluid acoustic structure coupling vibration is rare in nuclear power plant (NPP), especially in the stagnant branch of reactor coolant loop (RCL). This paper reports the phenomenon of fluid acoustic solid coupling vibration of one stagnant branch pipeline founding in the hot function test (HFT) of NPP. Field measurement shows that the vibration exceeds the limit value so it is necessary to determine the root cause of pipeline vibration and provide the modified layout. Based on the data of single point measurement, on-line monitoring, knocking test, acoustic modal calculation and modal analysis, the authors obtain the vibration characteristics and laws of pipelines and infers the acoustic solid coupling. The computational fluid dynamics (CFD) analysis and Strouhal number calculation, the inference of fluid acoustic coupling is identified. According to the above inference, the modified layout and the vibration prediction of the modified layout are proposed. After modification the HFT and the power up test was started. The vibration trend was consistent with the prediction, which verified that the fluid acoustic structure coupling was the root cause of the stagnant branch vibration. The amplitude of the modified the stagnant branch pipeline reduced in the whole process of temperature rise. The operation meets the limit requirements. The stagnant branch can operate normally.

Activated Corrosion Products Evaluations for Occupational Dose Mitigation in Nuclear Fusion Facilities

Activated corrosion products generation in primary heat transfer systems of nuclear fusion facilities is a relevant radiological source term for occupation dose assessments. The formation of the Chalk River Undefined Deposit, already well known in nuclear fission power plants, represents a significant safety issue in fusion applications due to the intense high energy neutron fluences (about 14 MeV in Deuterium-Tritium operation). The activated corrosion products formation is a multi-physical problem. The combined synergy of activation, corrosion, dissolution, erosion, ejection, precipitation, and transport phenomena induces the contamination of coolant loop regions located outside the bio-shield, where scheduled worker operation might take place. The following manuscript shows how activated corrosion products are evaluated for the nuclear fusion power plant design under investigation by the Safety and Environment Work Package (WPSAE) of the Eurofusion Consortium (i.e., the European Demonstration power plant, EU-DEMO). The major issues in activated corrosion products estimations are here exposed and the main results for mass and activity inventories are briefly shown for some main Primary Heat Transfer Systems of EU-DEMO.

On the validation of ATHLET 3-D features for the simulation of multidimensional flows in horizontal geometries under single-phase subcooled conditions

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

Методика поверочных расчетов параметров замкнутых газотурбинных установок космического назначения

The paper considers the methodology for checking calculations of parameters of space purpose closed-loop gas turbine units in various operating modes and changeable external conditions. The methodology is based on an algorithm for finding the conditions for the turbine-compressor joint operation in the energy conversion closed loop with subsequent clarification of the temperature of the working fluid at the inlet to the compressor. Finding the common operating point of the turbine and compressor is possible through using their universal flow-pressure characteristics obtained experimentally or calculated by the similarity methods of bladed machines. To clarify the temperature of the working fluid in front of the compressor, we also considered the conditions for ensuring a balance between the thermal power removed to the radiant cooler loop and the thermal power discharged by the radiant cooler into environment. The study briefly describes the software which implements the methodology and introduces the results of test calculations.

Development of an optimized thermodynamic model for VVER-1200 reactor-based nuclear power plants using genetic algorithm

In this work, an optimized thermodynamic modelforthe secondary coolant loop of a VVER-1200 reactor-based nuclear power plant isdeveloped. The values of some of the thermo-physical parameters such as coolant flow rate, coolant inlet and outlet temperatures from the steam generator and the turbines, etc. are obtained from the available literature. On the other hand, the unknown plant parameters such as feedwater heater pressures are predicted using a Genetic Algorithm-based optimization process. Three crossover rates, 0.5, 0.7 and 0.9, and three mutation rates, 0.01, 0.05 and 0.1, are considered. Results reveal that the value of the optimization process of the thermodynamic model is not significantly affected by the change in the crossover rate or mutation rate. The optimized model is compared with two theoretical models, a model available in the literature and experimental data. Results reveal that the predicted values of the optimized model are more accurate compared to the other available options. The optimized model also predicts that the maximum overall efficiency and the maximum output power of a plant situated in tropical region should be around 35.87% and 1152.16MWe respectively for 28 °C tertiary coolant temperature, down by 3.94% from the rated values in cold weather. However, the power output at 35 °C tertiary coolant temperature is calculated to be 1117.88 MWe, almost 6.69% lower than the rated value for VVER-1200. Thus, the efficiency of the plant is highly sensitive to the weather variation of a region.

Dynamic mass accountancy modeling of a molten salt reactor using equilibrium thermodynamics

A mechanistic-based mass accountancy model in the context of liquid-fueled molten salt reactors was implemented in the dynamic systems modeling software library TRANSFORM by way of coupling with the equilibrium thermodynamics code Thermochimica. Liquid-fueled molten salt reactors present new challenges for mass accountancy because of the dissolution of fuel and evolved fission products, which may be soluble in the salt, off-gas, or precipitate. Two cases of mass loss from the molten salt were addressed: off-gassing and precipitation. The software implementation was tested through a series of increasingly complex demonstration problems, culminating in a model of the primary fuel and primary coolant loops of the molten salt demonstration reactor. Analysis shows that negligible mass was lost from the salt under normal operating conditions, but an overheating event caused by partial loss of fuel loop cooling resulted in release of measurable amounts of uranium (among other elements) via off-gassing. The tools developed here are primarily aimed at capability development but are readily available for use in further modeling of molten salt reactor concepts. These tools have not yet been validated, and future experimental work to perform this validation is recommended.

ATHLET Simulation of PKL I2.2 IB-LOCA Benchmark Test and Quantitative Assessment

The Primary Coolant Loop Test Facility [Primӓrkreislӓufe Versuchsanlage (PKL)] PKL I2.2 Benchmark experiment for an intermediate-break loss-of-coolant accident (IB-LOCA) with a 13% or 17% break at the cold leg was performed in the Organisation for Economic Co-operation and Development/PKL-4 project at PKL in Erlangen, Germany. Analysis of Thermal-Hydraulics of LEaks and Transients (ATHLET) 3.1A was used at Karlsruhe Institute of Technology for its posttest calculations. Crucial predicted parameters were compared with measured data. The calculated trend of the selected parameters fits well with that of the experimental data except for the phenomenon of core heatup and the value of the peak cladding temperature. A fast Fourier transform–based method was chosen to quantify the matching of the parameter trends. According to the quantitative assessment, the IB-LOCA scenario and its detailed phenomena can be predicted well by ATHLET. Additionally, some discrepancies, i.e., insufficient reliable predictions for break mass flow and for reactor pressure vessel collapsed water level, were also observed, possibly deserving another study to undergo deeper scrutiny.

Experimental and Numerical Study on Pump-Driven Two-Phase Cooling Loop for High Heat Flux Avionics

Experimental and Numerical Study on Pump-Driven Two-Phase Cooling Loop for High Heat Flux Avionics

The probability of the existence of defects that lead to the destruction of the pressure vessel without leak

Relevance. To ensure the safety of a nuclear power plant on the basis of the requirements of norms and rules in the field of the use of atomic energy for pipelines of the primary circuit of a nuclear reactor, the design should apply the leak before break concept. The main idea of the concept is to prevent a sudden rupture of the pipelines of the reactor coolant loop, and consists in substantiating the fact that the rupture is preceded by the formation of a stable through crack, which is detected by the provided leak control means. When substantiating the concept, it is assumed that break without leak is an impossible event. This article provides a method for determining the probability of a failure event without a leak. Purpose - estimate the probability of the existence of a defect that can lead to the destruction of the vessel or pressure pipeline without leakage, as well as the probability of failure without leakage for a known number of loading cycles. Methods. To systematize the data obtained by different methods of non-destructive testing, conservative assumptions were used to determine the area of detected defects. On the basis of the obtained defect sizes, the defect size regions were determined, which can determine the scenarios of crack growth. Using the methods of mathematical statistics, the probability of the existence of a defect, which can lead to failure without leakage, was determined. Based on the methods of the theory of reliability, a comparison of the obtained probability of destruction with the admissible value is carried out. Results. A method for processing non-destructive testing data based on an assessment of the area of detected defects has been developed to systematize the data obtained by different non-destructive testing methods. The criterion for the development of cracks according to the scheme leak before destruction is determined. A method has been developed for determining the probability of a defect that can lead to failure without leakage. An example of calculation based on feed water pipelines is considered.

A Real-Time Simulator for an Innovative Hybrid Thermal Management System Based on Experimental Verification

The performance and efficiency of green energy sources in electric vehicles (EVs) are significantly affected by operation temperatures. To maintain the optimal temperatures of a hybrid energy system (HES), an innovative hybrid thermal management system (IHTMS) was designed. The IHTMS contains a coolant pump, a heat exchanger, a proportional valve for hybrid flow rates, five coolant pipes, and three electromagnetic valves to form two mode-switch coolant loops. A Matlab/Simulink-based simulator of the IHTMS was constructed by formulating a set of first-ordered dynamics of temperatures of coolant pipes and energy bodies using the theories of Newton’s law of cooling and the lumped-parameter technique. Parameters were majorly derived by measured performance maps and data from the experimental platform of the IHTMS. To properly manage the optimal temperatures, four control modes were designed for inner-loop form and outer-loop form. For the experimental platform to verify the simulator, two power supplies generated the waste heat of dual energy sources calculated by the driving cycle and vehicle dynamics. Simulation results show that the temperatures were controlled at their optimal ranges by proper mode/loop switch. With the inner-loop mechanism, the rise time of optimal temperature decreased 27.4%. The average simulation-experiment temperature error of the battery was 0.898 °C; the average simulation-experiment temperature error of the PEMFC was 4.839 °C. The IHTMS will be integrated to a real HES in the future.

Electrochemical corrosion study on base metals used in nuclear power plants in the HyBRID process for chemical decontamination

Base metal corrosion forms a significant issue during the chemical decontamination of the primary coolant loop in nuclear power plants as it is directly related to the economic and safety viability of decommissioning. In this technical note, potentiodynamic evaluations of several base metals (304 stainless steel, SA106 Grade B carbon steel, and alloy 600) were performed to determine their corrosion behavior during the hydrazine (N2H4)-based reductive ion decontamination (HyBRID) process. The results suggested that N2H4 protected the surface of the base metals in the HyBRID solution, which is primarily composed of H2SO4. The corrosion resistance of the carbon steel was further improved through the addition of CuSO4 to the solution. The corrosion rate of carbon steel in the H2SO4-N2H4-CuSO4 solution was lower than that exhibited in an oxalic acid solution, a commonly used reaction medium during commercial decontamination processes. These results indicate the superiority of the HyBRID process with respect to the base metal stability.

Tritium transport code development based on COMSOL and MATLAB and its application to the WCCB blanket

Water cooled ceramic breeder (WCCB) blanket was proposed as one of the candidate blanket schemes of the Chinese Fusion Engineering Test Reactor (CFETR). In this paper, a code simulating tritium transport in WCCB blanket and its external loops has been developed, where zero and two dimensional models are integrated to get more reliable results. The purge gas flow in the breeder zone, heat transfer in the blanket and tritium isotopes mass transport are considered. The isotope exchange in the purge gas and coolant loop are also taken into account in the model. The two dimensional model of the WCCB blanket and zero dimensional model of external loops are integrated to calculate tritium related quantities. The amount of tritium inventory in different components, tritium losses, tritium extracted by the tritium extraction system (TES) and purified by the coolant purification system (CPS) are obtained. The sensitivity analysis about key parameters has also been conducted.

A Computational Analysis of Heat Transfer in an Air-Cooled Proton Exchange Membrane Fuel Cell for Unsteady Flow Conditions Using a Reynolds Stress Turbulence Model

Within telecom applications, the air-cooled fuel cell, as a backup power system, has recently been gaining popularity. These fuel cells are cost effective, simple, and additionally they do not require a secondary coolant loop. However, since airflow is used as coolant, the heat generated by the fuel cell limits the current density and thus the power density. A turbulence inducing grid is placed in front of the cathode channel to make the airflow turbulent and hence increase the convective cooling effect from the air. The turbulence inducing grid is placed 2.5mm from the cathode channel since previous studies have shown this distance to be most effective.In this study, a three dimensional CFD analysis on the effect of a transient boundary conditions on the cooling effect of the turbulence grid has been conducted. Special emphasis has been on using the comprehensive Reynolds Stress Model as turbulence model. Additional analysis has been made on the cooling effect of pulsating flow with a period of 10 seconds excluding the turbulence inducing grid. At an electrode heat flux of 2600 W/m2, the results indicate that the convective cooling effect increases when using transient boundary conditions compared to steady state boundary conditions. Furthermore, results show that the start-up time is seven minutes, and the time-averaged peak temperature in the transient case is two degrees Celsius below the peak temperature of the steady state case. The CFD analysis show an additional cooling effect due to accelerated-decelerated flow conditions. Figure 1

Experimental Demonstration of Pressure-Driven Flash Boiling for Transient Two-Phase Cooling

This work demonstrates transient cooling for high-heat-flux electronics using pressure-controlled flash boiling. Many advanced electronics packages demand rapid, high-heat flux removal under stringent spatial and temporal constraints such as limited lateral heat spreading area, sharp power delivery transients, and heterogeneous power dissipation among dies. Flash cooling achieved by rapid depressurization of liquid coolant is inherently transient and provides high initial peak cooling that can be used for adaptive cooling demanded by these advanced packages. A coolant loop thermal testbed has been constructed to perform rapid transient thermo-fluid experiments of flash cooling with methanol. This work provides insights into depressurization-based flash incipience, thermodynamic analysis of flash cooling, and the ability to provide transient cooling in two application settings. First, a vapor chamber able to adapt to high initial transients applicable for short-burst applications. Second, a co-designed convective chamber that supports simultaneous power delivery and heat extraction for high-performance wafer-scale systems. A prototype vapor chamber was tested to obtain results for overall cooling and flash incipience experiments. The results demonstrate the efficacy of flash cooling for transient thermal management and highlight directions for further research and necessary data to exploit this technique for future electronics applications.

Tailoring Mission Effectiveness and Efficiency of a Ground Vehicle Using Exergy-Based Model Predictive Control (MPC)

To ensure dominance over a multi-domain battlespace, energy and power utilization must be accurately characterized for the dissimilar operational conditions. Using MATLAB/Simulink in combination with multiple neural networks, we created a methodology which was simulated the energy dynamics of a ground vehicle in parallel to running predictive neural network (NN) based predictive algorithms to address two separate research questions: (1) can energy and exergy flow characterization be developed at a future point in time, and (2) can we use the predictive algorithms to extend the energy and exergy flow characterization and derive operational intelligence, used to inform our control based algorithms or provide optimized recommendations to a battlefield commander in real-time. Using our predictive algorithms we confirmed that the future energy and exergy flow characterizations could be generated using the NNs, which was validated through simulation using two separately created datasets, one for training and one for testing. We then used the NNs to implement a model predictive control (MPC) framework to flexibly operate the vehicles thermal coolant loop (TCL), subject to exergy destruction. In this way we could tailor the performance of the vehicle to accommodate a more mission effective solution or a less energy intensive solution. The MPC resulted in a more effective solution when compared to six other simulated conditions, which consumed less exergy than two of the six cases. Our results indicate that we can derive operational intelligence from the predictive algorithms and use it to inform a model predictive control (MPC) framework to reduce wasted energy and exergy destruction subject to the variable operating conditions.

Experimental Study on the Dynamic Heat Transfer Characteristics of a Mechanically Pumped Two-phase Cooling Loop

The present research designs a mechanically pumped cooling loop system and conducts an experimental investigation on the dynamic heat transfer characteristics of the system. The effects of the start-up heat load and heat load variation on the dynamic characteristics of the system are analyzed. The results indicate that during the start-up period, as the heat load rises, the temperature overshoot and duration time of adjustment decrease, while the pressure oscillation amplitude increases. During the regular operation period, as the heat load increases, the temperature at the evaporator outlet gradually increases until it approaches the saturation temperature and then remains constant. The pressure evolution at the evaporator outlet can be divided into four stages: steady state, drastic oscillation, damped oscillation, and slight oscillation.

Heat transport and chimney design in a typical MTR reactor during natural convection cooling regime

In the present work, three dimensional CFD analysis is presented for a typical MTR reactor coolant loop during natural convection cooling regime. Design optimization studies are carried out. A model consisting of the core, the chimney, and the cold legs is simulated. The research reactor core is simulated as a porous zone with a cosine shape heat distribution. The chimney dimensions are varied to optimize its design for best thermal-hydraulic performance under natural convection conditions. The study is carried out for chimney extension ratio in the range of 1.5 ≤ Lch/Lcore ≤ 7.2 and chimney opening (expansion) ratio in the range of 1 ≤ B/b ≤ 2.5. Contours plot of temperature at different values of chimney extension and opening ratios were illustrated. The Nusslet number, coolant mass flow rate, maximum reactor core temperature, coolant velocity and core pressure drop are correlated as a function of the extension and opening ratios. Comparison with the results of CONVEC and RELAP5 codes shows a good agreement. The recommended optimum design conditions are for Lch/Lcore greater than 3.5, and B/b = 1 for best thermal-hydraulic performance.

The role of β-Zr in a Zr-2.5Nb alloy during aqueous corrosion: A multi-technique study

The Zr–2.5Nb alloy is used as pressure tubes in Canadian Deuterium Uranium (CANDU) nuclear reactors, and the typical starting microstructure consists of α-Zr grains elongated in both transverse and longitudinal directions and thin layers of partially decomposed β-Zr lying between the α-Zr grains. In this study, we have used state-of-the-art microscopy techniques to characterise the long-term thermally decomposed β phase in this alloy, and the oxide scale formed on them in a reactor coolant loop with the aim of understanding the mechanisms underpinning the thermal decomposition behaviour at service temperatures and exploring the role of the decomposed β-Zr phase in controlling the microstructure and microchemistry of the zirconium oxide, and hence its influence on the general corrosion resistance of the alloy. We observe that these β-Zr layers are heavily decomposed even after the short stress stage at 400 °C at the end of the manufacturing cycle, with a closely packed array of β-Nb precipitates forming in an α-Zr matrix. We have shown that the oxidation of these bands is significantly slower than the surrounding α-Zr matrix and that zirconium oxide grains are re-nucleated under each band. We conclude that it is the combination of the Nb-rich remnants of the original β-Zr layers arising from the hot extrusion and drawing stages and this new dense oxide that offers a significant barrier to the oxidation front (and also to the penetration of hydrogenic species), so the characteristic layered microstructure arising from the original manufacturing process is very important in determining the overall oxidation behaviour.

Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis

In this paper, the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Water-Cooled Lithium-Lead (WCLL) Test Blanket Module (TBM) Water Cooling System (WCS) from Europe is presented. The system consists of two loops in series. This design feature allows the removal of heat from the TBM box avoiding at the same time the release of radionuclides into the ITER Component Cooling Water System (CCWS), that acts as WCLL Test Blanket System heat sink. For this purpose, the WCS primary loop deals with the direct heat removal from the ITER TBM and the secondary one implements physical separation between the contaminated primary loop coolant and the CCWS. The insertion of an economizer into the primary loop determines the characteristic “eight” shape of the circuit. This choice was done in order to reduce the temperature difference on the intermediate heat exchanger. Hairpin type and steam bubble are the technologies selected for heat exchangers and pressurizers, respectively. Pressure and temperature control systems are foreseen to limit excursions from rated values in normal operational states and abnormal transients.A computational activity was promoted to assess the WCLL-WCS conceptual design, using a modified version of the RELAP5 Mod3.3 system code. A detailed thermal-hydraulic model was developed on the basis of design outcomes. The nodalization scheme includes the TBM, the WCS, a portion of the CCWS and the lithium-lead circuit. The computational campaign involved both the normal operational state and selected abnormal transients. In all the scenarios simulated, the conceptual design has highlighted the capability of operating the system respecting all the thermal-hydraulic requirements. The abnormal transient selected and presented is the loss of flow in the CCWS (loss of heat sink). In these conditions, TBM cooling function has been verified, keeping standard control strategies without any external action.