Secondary Coolant Research Articles

A thermodynamic study of water-based nanosuspensions as secondary heat transfer fluids in refrigeration systems

The suitability of nanosuspensions as secondary heat transfer fluids in refrigeration systems has been assessed aiming at the external irreversibilities of the heat exchangers of the refrigeration cycle. The results suggested that, in the case where the heat exchangers are optimized to minimize the rate of entropy generation, water-based nanofluids do not perform as well as pure water. Nanofluids enthusiasts, however, might advocate that nanosuspensions would allow the refrigerator to operate with lower condensing and higher evaporating temperatures, thus increasing the system COP. The present paper is therefore aimed at assessing both the internal and external irreversibilities of the refrigeration cycle when nanofluids are used as secondary coolants. The analyses have pointed out that despite most of the nanofluid research and development has been focused on the transfer properties (viscosity, conductivity), the transport properties related with the thermal capacity (specific heat, density) showed to be the most influential factors on the total system COP. In the case of water-based nanofluids, the nanoparticles have twice the density but only a quarter of the specific heat of pure water, suggesting that such nanosuspensions are not likely to perform better than the base fluid in HVAC-R applications.

Hybrid fuzzy-PID control of a nuclear Cyber-Physical System working under varying environmental conditions

In real-time control of Cyber-Physical Systems (CPSs), physical process variables are monitored and processed by intelligent controllers for keeping the values of safety parameters between given thresholds. Environmental conditions can affect the system dynamics and also the controller function. This paper presents a hybrid fuzzy-PID (Proportional-Integral-Derivative) controller, which learns the optimal PID controller parameters and adapts them according to the environmental parameters and process variables values.The proposed intelligent controller is applied to respond to the changes and transients in the environmental inlet air temperature of the secondary loop of a Lead-Bismuth Eutectic eXperimental Accelerator Driven System (LBE-XADS), for keeping the secondary average coolant temperature in a range of values between two safety thresholds.

Preventing iron(ii) precipitation in aqueous systems using polyacrylic acid: some molecular insights.

We present molecular dynamics simulations of aqueous iron(ii) systems in the presence of polyacrylic acid (PAA) under the extreme conditions that take place in the secondary coolant circuit of a nuclear power plant. The aim of this work is to understand how the oligomer can prevent iron(ii) deposits, and to provide molecular interpretation. We show how, to this end, not only the complexant ability is necessary, but also the chain length compared to iron(ii) concentration. When the chain is long enough, a hyper-complexation phenomenon occurs that can explain the specific capacity of the polymer to prevent iron(ii) precipitation.

A new method for D2O to H2O leak detection and identification of leaky heat exchanger in PHWR by Cerenkov photon counting technique

In pressurized heavy water reactors, leaks from D2O primary coolant and moderator to H2O secondary coolant and other light-water systems in heat exchangers cannot be completely ruled out. High cost of D2O demands that its loss should be prevented to maximum extent possible. Traditionally D2O leak detection and identification of leaky heat exchanger is carried out by measurement of tritium activity in H2O. Since tritium emits low energy beta radiation, its concentration in H2O is measured by mixing it with liquid scintillation solution in a definite proportion in counting vial and counted in a Liquid Scintillation Analyzer. It is very sensitive method for leak detection, but identification of leaky heat exchanger is time consuming and may require low power operation or reactor shut down. In the new method, high energy beta emitting fission products, which emit Cherenkov photons in H2O, were used as the tracer. H2O was poured in 20 mL plastic vials without scintillator and counted on Liquid Scintillation Analyzer. D2O leak was identified by comparing the Cherenkov photon count rate with that of the blank. A discrimination ratio significantly higher than average Cherenkov photon count rate for all heat exchangers was used to identify the leaky one. The technique has advantageous over existing method of D2O leak detection, such as, (1) scintillation chemicals are not required (2) low power operation or reactor shut down is not required for identifying the leaky heat exchanger (3) no generation of radioactive chemical waste (4) on-power leak identification reduces generation of radioactive liquid waste.

Design chemistry implementation experience during the power unit start-up and commissioning

The article covers the results of the design chemistry implementation during the commissioning of the innovative Unit 1 at the Novovoronezh NPP II equipped with a VVER-1200 reactor. The design chemistry is composed of the requirements for the primary and secondary coolant quality, recirculating coolant water (including essential service water), solutions used in safety systems as well as the requirements for technological tools maintaining their quality. Water chemistry setup operations play a significant role at all the stages of commissioning and low power testing. An analysis is made of essential system reactivation and cleansing stages, preliminary treatment technologies, primary and secondary circuit chemical water treatment, and radioactive water treatment. Some design advantages are highlighted, such as the use of reverse osmosis as one of the stages of water treatment and high-pressure filters on the bypass blowdown cleanup system. Consideration is given to some problematic issues that arose in the course of the start-up operations during the equipment depreservation, radioactive drain water treatment and in the recycling water supply system. The authors also analyze the design flaws and issues that may arise during long-lasting operation and the ways to solve them: (1) to provide a reference technology and equipment for processing radioactive drain water; (2) to exclude flushing with chlorinated hydrocarbons from the technology for depreserving the TG internal surfaces; (3) to apply water treatment with inhibitors providing calcium transport with a value close to 100% to the circulating water supply system with a cooling tower or provide liming of all additional water for southern NPPs to minimize the carbonate index.

<b>Energetic analysis of a commercial absorption refrigeration unit using an ammonia-water mixture

The ROBUR ® absorption refrigeration system (ARS), model ACF60, with a capacity of 17.5 kW, is tested, modeled and simulated in the steady state. To simulate the thermal load a heating system with secondary coolant was used, in which a programmable logic controller (PLC) kept the inlet temperature EVA at around 285.15 K. The mathematical model used was based on balancing the mass, energy and ammonia concentrations and completed by closing equations such as, Newton's cooling equation. The mathematical model was implemented using the Engineering Equation Solver – EES ® . The results obtained after modeling and a numerical permanent simulation are studied using the Duhring diagram. Potential points of internal heat recovery are visualized, and by using graphs of the binary mixture, it is possible to identify the thermodynamic states of all monitored points. The data obtained in the numerical simulation of the ARS was compared with data acquired in the actual tests of the ARS with the ROBUR ® apparatus.

Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

Numerical Study on the Secondary Air Performance of the Film Holes for the Combined Impingement and Film Cooled First Stage of High Pressure Gas Turbine Nozzle Guide Vane

Abstract The present numerical investigation of Leading Edge (LE) Nozzle Guide Vane (NGV) is considered with five rows of impingement holes combined with five rows of film cooled for the secondary coolant flow path analysis. The coolant mass flow rate variations in all the LE rows of the film holes externally subjected to the hot main stream were obtained by making a three-dimensional computational analysis of NGV with a staggered array of film cooled rows. The experiments were carried out for the same NGV using Particle Image Velocimetry technique to determine the effused coolant jet exit velocity at the stagnation row of film holes as mentioned in reference [Kukutla PR, Prasad BVSSS. Secondary flow visualization on stagnation row of a combined impingement and film cooled high pressure gas turbine nozzle guide vane using PIV technique, J Visualization, 2017; DOI: 10.1007/s12650-017-0434-6]. In this paper, results are presented for three different mass flow rates ranges from 0.0037 kg/s to 0.0075 kg/s supplied at the Front Impingement Tube (FIT) plenum. And the mainstream velocity 6 m/s was maintained for all the three coolant mass flow rates. The secondary coolant flow distribution was performed from SH1 to SH5 row of film holes. Each row of a showerhead film hole exit coolant mass flow rate varied in proportion to the amount of coolant mass rates supplied at the FIT cooling channel. The corresponding minimum and maximum values and their film hole locations were altered. The same behaviour was continued for the coolant pressure drop and temperature rise from SH1 to SH5 row of film holes. Owing to the interaction between hot main stream and the coolant that effuses out of the film holes, occasional presence of hot gas ingestion was noticed for certain flow rates. This caused nonlinear distribution in mass flow, pressure drop and temperature rise. The minimum flow rate results estimate oxidation of NGV material near the film cooled hole. And the effect of hot gas ingestion on the ejected film cooled jet which would recommends effective oxidation resistant material which in turn leads to better durability of the NGV surface.

Laboratory tests on heat treatment of ballast water using engine waste heat

ABSTRACTWaste heat recovery from shipboard machineries could be a potential source for heat treatment of ballast water. Similar to a shipboard schematic arrangement, a laboratory-scale engine-heat exchanger set-up harvesting waste heat from jacket water and exhaust gases was erected to test the level of species’ mortalities. Mortalities were also assessed under experimental conditions for cultured and natural plankton communities at laboratory level. Effect of pump impellers on species’ mortalities were also tested. Exposures between 60°C and 70°C for 60 sec resulted in 80–100% mortalities. Mortalities due to pump impeller effects were observed in the range of 70–100% for zooplankton. On the laboratory-scale arrangement, >95% mortalities of phytoplankton, zooplankton and bacteria were recorded. It was demonstrated that the temperature of tropical sea waters used as secondary coolant can be raised to cause species’ mortalities, employing engine exhaust gases. The results also indicated that pump impeller effects will enhance species’ mortalities. The limitations of the shipboard application of this method would be the large ballast volumes, flow rates and time for treatment.

The coolant purification system in DEMO: Interfaces and requirements

The blanket concepts investigated under the EUROfusion program rely on water or helium as primary or secondary coolant medium; main duty of the coolant is to recover the thermal power from the first wall and the blanket units and drive it into the Primary Heat Transfer System (PHTS). The coolant path goes through three different systems: the breeder, the tritium plant and the PHTS. In the breeder region, part of the produced tritium can permeate into the coolant loop, from the coolant into the steam generator and thus reach the environment. The Coolant Purification System (CPS), located inside the tritium plant, is responsible to recover tritium from the coolant and to control the coolant chemistry.The paper provides the status of the CPS design: the technology review was completed last year, while now the CPS interfaces and requirements are defined. Main issues arise from the definition of the amount of coolant to be routed inside the CPS due to the large uncertainties related to the amount of tritium permeation rate from blanket into coolant and to the presence and effectiveness of anti-permeation barriers.

Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6–12 to 2.0–2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains рН25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

Source Term Study on Tritium in HTR-PM: Theoretical Calculations and Experimental Design

The high temperature gas-cooled reactor pebble-bed module (HTR-PM) in China received much attention for its inherent safety performance and high thermal efficiency. The generation mechanism, distribution, reduction route, and release type of tritium (H-3) in HTR-PM are presented with a complete theoretical model. The calculation results indicate the majority of H-3 in the core is generated by the activation reaction of B-10. The activity concentration of H-3 in the primary loop and the specific activity of H-3 in the secondary loop at the operating equilibrium are computed as 3.69 × 106 Bq/m3STPof helium and 4.22 × 104 Bq/kg of water, respectively. The H-3 sampling measurement in HTR-PM has been designed to collect data from the primary coolant, from the liquid waste storage tank, from the secondary coolant, and from the liquid and gaseous effluents, separately. In this paper, the design of H-3 sampling positions in the helium purification system is discussed. The H-3 sampling measurement from the primary helium in HTR-PM has been improved, which can provide reliable activity concentration data of H-3 in the primary loop and supply accurate evaluation for the efficiency of the helium purification system.

Study of secondary flow and coolant film interaction in a high subsonic cascade

Secondary flows can significantly alter the surface flow field in a blade or vane passage of a turbomachine. The relationship between the three-dimensional flow field and the surface flow is essential to understand for performance improvement in gas turbines, where a secondary air flow (coolant) is injected through discrete holes provided on the endwall for cooling. The goal of the present investigation is to quantify the impact of secondary flows on film cooling and vice versa using surface flow visualization, film cooling effectiveness, and film cooling hole discharge coefficients. A high subsonic cascade with an isentropic throat Mach number of 0.68 was used for the experiments. Discrete film injection was provided on the inner endwall using a single row of cylindrical film cooling holes. It was observed that the saddle point shifted towards the airfoil leading-edge with an increase in film cooling blowing ratio. Streamlines showed that the leading-edge horseshoe vortex was confined to approximately 2% of the airfoil span for the no blowing case and intensified with film injection. Increasing the blowing ratio improved the film cooling effectiveness on the endwall. The film cooling hole discharge coefficients showed a maximum variation at the lowest blowing ratio and exhibited strong dependence on the local mainstream static pressure and the location of separation lines.

Experimental Study of DRACS Thermal Performance in a Low-Temperature Test Facility

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride salt–cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS heat exchanger (DHX) and the natural draft heat exchanger (NDHX). To experimentally investigate the thermal performance of the DRACS, a scaled-down low-temperature DRACS test facility (LTDF) has been constructed. The design of the LTDF is obtained through a detailed scaling analysis based on a 200-kW prototypic DRACS design developed at The Ohio State University. The LTDF has a nominal power capacity of 6 kW. It employs water pressurized at 1.0 MPa as the primary coolant, water near the atmospheric pressure as the secondary coolant, and ambient air as the ultimate heat sink. Three accident scenarios simulated in the LTDF are discussed in this paper. In the first scenario, startup of the DRACS system from a cold state is simulated with no initial primary coolant flow. In the second scenario, a reactor coolant pump trip process is studied, during which a flow reversal phenomenon in the DRACS primary loop occurs. In the third scenario, the pump trip process is studied with a simulated intermediate heat exchanger in operation during the simulated core normal operation. In all scenarios, natural circulation flows are developed as the transients approach their quasi steady states, demonstrating the functionality of the DRACS. The accident scenarios in the prototypic FHR design corresponding to the simulated ones in the LTDF are also predicted by following a scaling-up process. The predictions show that at any time during the simulated transient, the salt temperatures will be higher than their melting temperatures and that therefore there will be no issue of salt freezing in the three projected accident scenarios. However, the scaled-up primary salt temperatures indicate that the prototypic DHX may have been undersized and may need to be redesigned.

원전 2차계통수 모사 환경에서 용접배관 감육 특성에 미치는 재료 및 유속의 영향

The pipelines and equipments are degraded by flow-accelerated corrosion (FAC), and a large-scale test facility was constructed for simulate the FAC phenomena in secondary coolant environment of PWR type nuclear power plants. Using this facility, FAC test was performed on weld pipe (carbon steel and low alloy steel) at the conditions of high velocity flow (> 10 m/s). Wall thickness was measured by high temperature ultrasonic monitoring systems (four-channel buffer rod type and waveguide type) during test period and room temperature manual ultrasonic method before and after test period. This work deals with the complex effects of flow velocity on the wall thinning in weld pipe and the test results showed that the higher flow velocity induced different increasement of wall thinning rate for the carbon steel and low alloy steel pipe.

Heat transfer in a two-inlet rotating wedge-shaped channel with various locations of the second inlet

The heat transfer characteristics of a two-inlet wedge-shaped channel with lateral fluid extraction are experimentally investigated under both rotating and non-rotating conditions. Three different locations of the second inlet are tested. The effective inlet Reynolds number is fixed at 15,000 whereas the mass flow rate ratio (MR, second inlet mass flow rate/major inlet mass flow rate) ranges from 0 to 1.0. The highest rotation number is around 0.23. The results show that injecting the secondary coolant at the top of the channel should be the best option in the view of optimizing the bulk temperature and the heat transfer of the high-radius corner of the channel in both rotating and non-rotating cases. Two critical MR can be identified in the non-rotating channels. Higher the radius of the coolant injection, lower the critical MR. The magnitude of the second critical MR is double of the first one in all three cases. However, the local critical mass flow ratio (MRx) is almost independent of the location of the second inlet. In both rotating and non-rotating cases, operating at high MR over the critical point is not recommended because the cooling air is not used efficiently. The influence of the second coolant injection is limited in the vicinity of the major inlet, which brings the convenience to optimize the local heat transfers without disturbing the heat transfer at the upstream locations.

Evaluation Of Operating Performance Of The Reactor Coolant System Of Rsg-Gas Using Chemcad6.1.4

<p>Evaluation of the operation of reactor main coolant system RSG-GAS is very important. This is related with the operation of the reactor that has been going on more than 27 years although the current reactor coolant system is still operating safely. The main coolant system RSG-GAS including both the primary and secondary cooling system. Evaluation can be conducted by calculating the coolant system performance by using the ChemCad6.1.4 software. Furthermore, this calculation result is compared with the most recent measurement in the cooling system operation. By this way it can be known the performance degradation in the Reactor Coolant System operation.The operating parameter data of primary and secondary cooling system are temperature, pressure and flow rate of coolant. The reactor operation at the power of 15 MW is used as a reference data.The calculation result shows that the temperature parameter between calculation by ChemCad and the measurement data qualitatively is no significant deviations. However in the secondary coolant system, the outlet pressure of both heat exchanger nozzles are 0.8bar and 1.2bar that indicate lower than the calculation results. It’s may be caused by the flow disturbances through the tube-side of heat exchangers. In generally the existing operating parameters on the reactor cooling system in terms of safety aspects still under safely condition.</p>

Erratum to: “Development and application of an information–analytic system on the problem of flow accelerated corrosion of pipeline elements in the secondary coolant circuit of VVER-440-based power units at the Novovoronezh Nuclear Power Plant”

Erratum to: “Development and application of an information–analytic system on the problem of flow accelerated corrosion of pipeline elements in the secondary coolant circuit of VVER-440-based power units at the Novovoronezh Nuclear Power Plant”

Key Findings from the Artist Project on Aerosol Retention in a Dry Steam Generator

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generators (SGs) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

Multiscale integral analysis of a HT leakage in a fusion nuclear power plant

The present work presents an example of the application of an integral methodology based on a multiscale analysis that covers the whole tritium cycle within a nuclear fusion power plant, from a micro scale, analyzing key components where tritium is leaked through permeation, to a macro scale, considering its atmospheric transport. A leakage from the Nuclear Power Plants, (NPP) primary to the secondary side of a heat exchanger (HEX) is considered for the present example. Both primary and secondary loop coolants are assumed to be He. Leakage is placed inside the HEX, leaking tritium in elementary tritium (HT) form to the secondary loop where it permeates through the piping structural material to the exterior. The Heating Ventilation and Air Conditioning (HVAC) system removes the leaked tritium towards the NPP exhaust. The HEX is modelled with system codes and coupled to Computational Fluid Dynamic (CFD) to account for tritium dispersion inside the nuclear power plants buildings and in site environment. Finally, tritium dispersion is calculated with an atmospheric transport code and a dosimetry analysis is carried out. Results show how the implemented methodology is capable of assessing the impact of tritium from the microscale to the atmospheric scale including the dosimetric aspect.