Heat Removal System Research Articles

Research on design requirements for passive residual heat removal system of lead cooled fast reactor via model-based system engineering

Traditional text-based system engineering, which has been used in the design and application of passive residual heat removal system (PRHRS) for lead-cooled fast reactors, is prone to several problems such as low development efficiency, long iteration cycles, and model ambiguity. This study aims to effectively address the above-mentioned problems by adopting a model-based system engineering (MBSE) method, which has been preliminarily applied to meet the design requirements of a PRHRS. The design process has been implemented based on the preliminary design of the system architecture and consists of three stages: top-level requirement analysis, functional requirements analysis, and design requirements synthesis. The results of the top-level requirements analysis and the corresponding use case diagram can determine the requirements, top-level use cases, and scenario flow of the system. During the functional requirements analysis, the sequence, activity, and state machine diagrams are used to develop the system function model and provide early confirmation. By comparing these sequence diagrams, the requirements for omissions and inconsistencies can be effectively checked. In the design requirements synthesis stage, the Analytic Hierarchy Process is used to conduct preliminary trade-off calculations on the system architecture, after which a white box model is established during the system architecture design. Through these two steps, the analysis and design of the system architecture are ultimately achieved. The resulting system architecture ensures the consistency of the design requirements. Ultimately, a functional hazard analysis was conducted for a specific incident to validate case requirements and further refine the system architecture. Future research can further reduce the design risk, improve the design efficiency, and provide a practical reference for the design and optimization of PRHRS in digital lead-cooled fast reactors.

Convolutional neural network-based pattern recognition in natural circulation instability images

Heat removal systems employing natural circulation are key in new nuclear power plant designs for mitigating accidents. This study applies Convolutional Neural Networks (CNNs) to classify 'chugging' instability phases, analyzing 1152 two-phase flow images from a Natural Circulation Circuit. Three CNN models, including one incorporating transfer learning from the ImageNet database, were trained via five-fold cross-validation to fine-tune hyperparameters. This involved comparing models with and without transfer learning against a baseline linear model. A model using a pre-trained Resnet50 with transfer learning accurately classified all 230 samples, outperforming the baseline linear model with an F1-Score of 0.859. The results endorse the use of CNNs with transfer learning for thermohydraulic image analysis in identifying natural circulation instability stages.

Transient simulation and analysis of a supercritical CO2 heat removal system under different abnormal operation conditions

The supercritical carbon dioxide (sCO2) heat removal system, which is based on multiple closed Brayton cycles with sCO2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this study considers different sudden failures during the accident progress, e.g. failure of single sCO2 cycles, control systems and valves. These abnormal conditions were investigated with the thermal-hydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up can cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible.

Multi-channel general reservation of NPP own needs on the basis of combination with autonomous hydrogen energy complex

The authors have developed an autonomous hydrogen energy complex and analyzed its economic efficiency. The developed energy complex is multifunctional. The main result of the work is the production and supply of electricity to the power system during hours of increased load. In addition, the low-power steam turbines included in the energy complex can provide plant-wide backup of the NPP's own needs in the event of a station blackout: based on the experimental data of the Balakovo NPP, it was previously shown that one power unit with a low-power turbine using the energy of the residual heat from the reactor can for a long time to provide electricity to required power units. Thanks to the installation of low-power turbines as part of the hydrogen energy complex, their constant warm state and self-sufficiency will be ensured by generating electricity for the power grid or for own needs in the normal mode. The paper considers options for installing an autonomous hydrogen power complex with and without replacing expensive external heat exchangers of the system for passive removal of heat from the reactor core. The efficiency of an alternative option without installing a storage system with the sale of off-peak electricity to the power system, which, in the option with AHC, is used to generate hydrogen/oxygen by electrolysis, depending on the tariff for off-peak electricity, is also calculated. Due to the lack of existing analogues of the hydrogen energy complex, several options for the proposed investment were considered. The results of the study revealed that the replacement of expensive equipment of the passive heat removal system significantly increases the economic efficiency of the proposed redundancy system. For the accepted options for capital investments, the boundary values of the off-peak selling tariff for electricity are shown, at which the use of the storage system is more efficient in comparison with the sale of nightly electricity to the power system.

Study on the seawater cooled PRHRS of a nuclear propulsion ship with the upper heat exchanger for driving force and the seawater heat exchanger as an ultimate heat sink

As international maritime trade among countries increases, great efforts are being made to reduce the amount of greenhouse gases emitted from ships using conventional fossil fuels. A variety of environment-friendly power sources for ships are being considered, and nuclear energy is one of them. The research described in this paper focuses on elemental technologies related to nuclear-powered ships rather than the nuclear reactor technology itself. Among those technologies, the Passive Residual Heat Removal System (PRHRS) of a nuclear-powered ship is studied to remove residual heat by using seawater as an alternative heat sink. New concept of PRHRS is proposed and it is named as SWC-PRHRS. SWC-PRHRS has two heat exchangers: one is for driving force and another is for dumping the most of residual heat to the sea. Numerical analysis and experimental studies are conducted to improve the performance of SWC-PRHRS. The numerical results show that most of the residual heat comes from a reactor core is eliminated by sea water cooling proving the concept of SWC-PRHRS is working properly without any operator intervention and additional water supply. Experimental studies are conducted based on the numerical analysis results. From the experimental results, it is found that the mass flow rate of SWC-PRHRS is very sensitive to the filling ratio of SWC-PRHRS and there are three distinctive filling ratio ranges. Through the experiment, it is confirmed that SWC-PRHRS is working efficiently with the lower mass flow rate and the smaller upper heat exchanger without any operator intervention. SWC-PRHRS could remove a large amount of residual decay heat from the ship's reactor and release it to the sea. Considering the size, weight, and exceptional heat removal capability of SWC-PRHRS, it is clear that SWC-PRHRS is well suited for marine application usage.

Experimental study of the natural deposition of submicron aerosols on the surface of a vertical circular tube with non-condensable gases

This paper constructed an experimental platform containing a space heat exchanger tube, and performed aerosol deposition experiments under different deposition mechanisms. The objective was to study the effect of the passive containment heat removal system on the natural deposition of submicron aerosols during post-accident operation of severe accidents. By controlling the variables, we obtain the impact of gravity deposition, thermophoretic, and diffusiophoresis deposition mechanisms on submicron aerosols in the presence of a space heat exchanger tube using the experimental data stripping method. The study found that the gravitational settling rate of aerosols in two-phase mixtures at the same temperature and pressure is about 1.5 times higher than in dry air. Although submicron aerosols are subject to thermophoretic settling in the presence of temperature differences, due to the small surface area of the heat exchanger tube, the potential thermophoretic sedimentation accelerating effect on the particles can be ignored if the mixed gas temperature difference is within 2000 K/m. Unlike the steam condensation on the flat plate surface, when vapor condenses on the heat exchanger tube surface, as the steam proportion and condensation rate increase, it is easy to cause the accumulation of non-condensable gas. This phenomenon can reduce the reverse diffusion slip of the particles and enhance the influence of Stefan flow in the diffusionphoresis deposition process. The above effects make the experimental results of aerosol diffusiophoresis deposition closer to the SF model when the vapor fraction is 70 %. When the vapor fraction is 50 %, the predictions of the WM model, which considers the average mass flow of the gas mixture, are more applicable. When the condensation rate continuously increases, the convective mass transfer of vapor condensation will be enhanced, and the deposition rate of particles will be further enhanced.

Application of REPAS to analyze the sump clogging issue following a LOCA and its impact on the reliability of the ECCS long-term core cooling function

Sump clogging has been identified as a relevant issue after an accident occurred in the Barsebäck-2 nuclear power plant in Sweden (1992). Following a steam line Loss of Coolant Accident (LOCA) due to the inadvertent opening of a safety relief valve, the jet stripped some insulation material from nearby pipes. The insulation debris were transported to the inlet of the strainers for the drywell spray system, thus clogging the intake. The accident was not serious but showed that the Emergency Core Cooling System (ECCS) could have failed. As a consequence, several actions have been undertaken by international organizations, regulatory bodies and nuclear power plant owners to characterize this issue, and propose solutions and improvements. In the present paper, the Reliability Evaluation of Passive Safety Systems (REPAS) methodology is applied to analyze the sump clogging issue and its effect on the long-term core cooling function. Originally developed to evaluate the reliability of passive systems, REPAS is here applied for the first time to an active system. The application is performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code, developed by the USNRC, to simulate a generic three-loop PWR, with an active decay heat removal system.

ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate-Type Heat Exchanger

This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E-SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate-type one coupling the primary side to the secondary side, and a vertical tube one coupling the secondary side to the water pool. Such a system is considered to be the most effective passive system, capable of safely managing the SMR accident and accidental situations, and achieving long-term decay heat removal without the need for electricity or external inputs. A description of the primary and secondary loops of the plant is given, together with the installed instrumentation and data acquisition system. In addition, the paper summarises the tests performed in terms of test procedures, test type and associated objectives, test matrix, test results, achievements, and open issues.

Qualification of Compensatory Test Methods Involving Pressure Increase in the VVER Containment System

In general, containment leakage testing of nuclear power plants with VVER reactors by elevated pressures using the compensatory leakage detection method (CLDM) can be performed at pressures and pressure increase rates higher than those stipulated by the regulations using the absolute pressure method (AP method). Elevated pressures and pressure increase rates under certain conditions can violate safety limits of the containment systems and/or decrease the reliability/life of containment structures and equipment. These factors determine the need to qualify the CLDM to promote conditions for reliability and safety. A non-stationary thermodynamic model of the qualification conditions for CLDM testing of the containment was developed. The criteria for CLDM qualification conditions are the maximum allowable pressure and pressure increase rate during testing. The CLDM condition for recording leakage in the containment is pressure stabilization in the containment systems. Based on the developed CLDM thermodynamic model, it was established that the containment leakage rate is determined by the flow rate of air entering the containment systems and by the thermodynamic state of the air inside and outside the containment systems. The established qualification conditions were used to determine conditions for the minimum recorded leakage sizes within CLDM and the maximum allowable ambient air flow rates and test duration. A prerequisite for justifying the qualification and implementation of the CLDM is to revise/amend regulatory and technical requirements for the maximum allowable pressure and pressure change rate in the containment systems and for the conditions for disconnecting containment systems passive heat removal systems (if any) during testing.

Neutronic analysis of a heat-pipe cooled reactor

Heat-pipe cooled reactors belong to the group of nuclear reactors using heat pipes filled with liquid metals (such as sodium or potassium) to cool the core. Due to the passive heat removal system, there is no need to use closed loops with pumps, and the reactor can be operated with reduced operational requirements. Consequently, this system can be used in remote locations without access to an electrical grid, or it can be used for space applications.This paper deals with a neutronic study of the Special Purpose Reactor Design-B concept from the Idaho National Laboratory. Using Monte Carlo code Serpent 2 and ENDF/B VIII.0 library, a fixed-temperature model was created to calculate the safety characteristics of the system. This included reactivity coefficients, power distribution, neutron flux spectrum, and criticality safety. In a simplified depletion calculation, an effect of fuel depletion on safety systems was determined as well as a decay heat.

Preliminary analysis of maximum hypothetical accident for a solid core space nuclear reactor power system

In order to study the safety characteristics of the solid core space nuclear reactor power (SNRP) system under the maximum hypothetical accident, a two-dimensional entire core transient heat transfer analysis model was established, and the key parameters response characteristics of the ultra-small lithium-cooled SNRP under two maximum hypothetical accidents, namely, loss of heat sink (LOHS) and loss of coolant accident (LOCA), were calculated and analyzed. In the maximum hypothetical accidents, the coolant cooling capacity is lost, thus the core decay power is discharged into space only through the radiation of the residual heat removal system and the side surface of the reactor vessel. During the accidents, the heat transfer of the core deteriorated, and the fuel temperature may rise to the melting point, resulting in radioactive leakage. The results show that: (1) In the LOHS accident, the maximum fuel temperature reaches 2150 K at 550 s, and the pressure of the primary system volume accumulator continues to increase to the set system pressure safety limit of 2 MPa, resulting the primary loop overpressure failure. And the fuel matrix temperature is close to the set cladding limit temperature of 2200 K; (2) In the LOCA, the deterioration of heat transfer in the core makes the temperature increase rapidly, reaching a maximum of 3016 K at about 630 s, which is very close to the melting temperature of UN fuel 3123 K. As the decay power decreases, the maximum core temperature decreases to less than 1600 K after 24 h of the accident. The auxiliary cooling system of the solid core SNRP system under the maximum hypothetical accident is optimized, and the design parameters of the auxiliary cooling system meeting the safety requirements are obtained.

Preliminary design of the suppressive containment system based on HPR1000

The current HPR1000 has a high construction cost and weak economic competitiveness due to its complex system and large containment. This study proposes a conceptual design of an advanced containment system based on HPR1000, which aims at reducing the construction cost of nuclear power plants without reducing their safety. In the design, the In-containment Refueling Water Storage Tank (IRWST) and spray system were discarded, and the space of the double containment vessel was used to create an annular cavity pool (ACP). The transient pressurization in the containment caused in the early stages of the accident can be quickly suppressed by ACP instead of the traditional large-volume suppression concept. And due to the significant condensation of steam within the ACP water, the transportation of air into the ACP results in a reduction in the air mass fraction in the vicinity of the passive containment heat removal system (PCHR) located within the containment structure. This phenomenon subsequently leads to an enhancement in heat transfer performance. In addition, ACP also serves as the water source for the safety injection system, the reactor cavity water injection system, and the core replacement system. Compared with HPR1000, the system of the preliminary design has been greatly simplified, and the size of the containment vessel has been reduced by nearly 47%. The simulation of the large-break loss-of-coolant accident (LBLOCA) scenario has been conducted to examine the response of the containment system, specifically focusing on the performance of the safety injection system and suppression system. The results show that the integrity of the containment and core was ensured throughout the accident, and there was a safety margin.

Experimental and numerical analysis of the CO shift process for purified hydrogen production in an integrated gasification fuel cell

The Korean government has recently begun to focus on securing adequate hydrogen production to meet demands for the development and disseminaetion of sustainable technologies. This study describes a Korean national project developed for this hydrogen production purpose. In this study, we investigated the CO shift process for producing purified hydrogen in an integrated gasification fuel cell using experimental and numerical analyses. First, we successively designed and constructed a system with a syngas input of 160 Nm3/h; this system was efficiently operated for 45 h and we thus classified the operation into five spans (Start-Up, Heat-Up, Syngas Feeding, Steady-State, and Shut-Down). For example, in the span of State-State, the syngas load from 160 to 200 Nm3/h under the same H2O feeding conditions. Second, based on a computational flow dynamic model and its verification (comparison with experimental results), we conducted a parameter study on the effects that water injection has on the high-temperature CO shift reactor and showed that cases with a well-designed heat removal system yielded higher CO than other cases. The results of this study can be used as a guideline to develop CO shift processes at commercial grades (i.e., 10,000 Nm3/h syngas) for the production of purified hydrogen.

Thermal-hydraulic modelling and stade staty analysis of a LS-VHTR reactor

The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) is a liquid salt cooled and TRISO fueled reactor, which presents very good features in terms of power production and safety aspects. The advantage of using liquid salt as coolant is mainly related with its high efficiency of heat transfer and operation at low pressures. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core, in steady state conditions, starting from the single channel level and concluding with the full core model. The LS-VHTR is a new reactor concept that introduces some different features in relation to other nuclear reactors. In this way, a thermal modeling of this reactor has been developed using the RELAP5-3D code using a point kinetic model. A modeling with three channels has been presented. The core coolant temperature increase, the core coolant mass flow and the pressure along the core were simulated and presented in good agreement with the expected behavior

Autonomous Thermosiphon System of Passive Residual Heat Removal from the Primary Circuit of the Reactor Plant: Features of Operation, Characteristics and Basic Advantages

An autonomous system of passive removal of residual heat (PRRHS) of a reactor installation with VVER designed to ensure the safety of nuclear power plants in an accident with complete long-term blackout is considered. The system provides for the removal of heat directly from the first circuit of the reactor plant (PRRHS R). In order to increase the reliability and safety of the emergency heat sink, heat exchange equipment based on closed-type evaporation and condensation devices – two-phase thermosyphons – has been used in the system. The main feature of such heat exchangers is that their thermosiphon assemblies structurally separate the primary circuit and the auxiliary circuit of the PRRHS, which is removed outside the reactor compartment, and provide safe and efficient heat removal, reduce the risk of radioactive contamination spreading beyond safety barriers. Such autonomous passive systems will provide effective heat removal directly from the primary circuit by changing the chain of successive heat transfer sites from nuclear fuel to the final absorber and excluding from it such elements, as for example steam generators, the condition and operability of which in the emergency process of heat removal have a major impact on the safety of the reactor core. The article presents a diagram of an autonomous heat sink system; also, a description of its operation is given. The main characteristics of the course of the emergency process of removal of residual heat by the autonomous thermosiphon PRRHS R obtained by computational modeling have been considered. The advantages of an autonomous thermosiphon passive system in comparison with a passive heat removal system of a reactor installation with VVER through the second circuit are analyzed. The obtained results are proposed to solve the problems of diversification of passive safety systems of evolutionary reactor plants of nuclear power plants with VVER type reactors.

Analysis of a parallel-channel natural circulation loop in the passive residual heat removal system of the NHR200-II reactor

The passive residual heat removal (PRHR) system of the NHR200-II nuclear heating reactor is a parallel-channel single-phase natural circulation loop. The reverse-flow phenomenon in parallel-channel natural circulation loops will cause insufficient heat removal capacity in the PRHR system. In this study, a simplified mathematical model was established for the PRHR system of NHR200-II. The mechanism of the reverse flow phenomenon in the PRHR system was analyzed using a pressure drop-flow rate diagram, and the effects of cooler elevation and loop resistance on the number of reverse flow branches were discussed. The analysis reveals that the reverse flow phenomenon of the PRHR system can be suppressed by increasing the elevation of the air cooler. Furthermore, there exists a critical elevation of the air cooler. It is not feasible to suppress the reverse flow by changing the friction coefficients of either the PHE branches or the air cooler branch when the air cooler elevation is less than the critical value. Oppositely, the number of reverse flow branches can be reduced or even eliminated by increasing the resistance of the PHE branches or reducing the resistance of the air cooler branch when the air cooler elevation is greater than the critical value.

Numerical simulation of the operational characteristics of air-cooled PRHR under SBO accidents

In this paper, the operational characteristics of air-cooled type passive residual heat removal (PRHR) system of the nuclear heating reactor are studied with RELAP5 code. The PRHR uses the air natural circulation which can transfer heat directly from the reactor vessel to the air, and its heat transfer process is particularly important for the reactor safety. However, the current thermal–hydraulic code lacks the relevant air cooler model. A method how to model an air cooler is proposed based on the conservation equation and can be widely applied to different situations. Moreover, the operating characteristics of air cooling tower and the transient thermal–hydraulic characteristics of station blackout (SBO) event have been analyzed. The results show that the heat transfer coefficient of air cooler increases with the heat transfer power, and the inlet temperature has little influence on the heat transfer coefficient. The air cooling type PRHRS can transfer the core residual heat timely and continuously. There is a time delay of the heat exchanger power between the main heat exchanger and air cooler caused by the temperature distribution.

PIV-based experiments on reverse flow characteristics in an equal-height-difference passive heat removal system for ocean nuclear power plants

The equal-height-difference passive heat removal system (EHDPHRS) has a large potential in floating nuclear power plants (FNPPs) due to its ability to fully utilize the marine environment to achieve residual heat removal. However, the horizontal connection between the heat sink and the heat source could easily lead to the subcooled-water reverses flow and thus trigger serious flow instabilities. In the present paper, a fully visualized experimental set was built, combined with a particle image velocity (PIV) system to investigate the subcooled-water reverses flow in the EHDPHRS. First, PIV observations clearly highlighted five flow stages during a complete natural circulation, including the reverse flow occurred in the single-phase and two-phase flow. Then, it was shown that the single-phase reverse velocity is proportional to heating power and inversely proportional to subcooled-water temperature. While, the two-phase reverse velocity increases first and then decreases with the increase in heating power. Moreover, the two-phase reverse flow phenomenon shows obvious periodicity. Finally, the buoyancy force (Fb) caused by the density contrast, and the pressure difference (FΔP) caused by the direct contact condensation (DCC) are the triggering forces for the reverse flow, while the flow inertia force (Fi) caused by the forward flow represents the restraining force.

The Passive Heat Removal Facility: Testing and Demonstrating an Innovative Decay Heat Removal System

Innovative reactor concepts are being studied by several research institutions and private entities for their role against climate change and energy poverty. The Generation IV International Forum, committed for more than 20 years in support of advanced reactors, has drawn up a series of objectives for the new class of nuclear plants, among which is an improvement in safety and economy through passive safety systems for the removal of decay heat. One of the most studied technologies is the lead fast reactor (LFR), whose coolant has a high boiling point, excellent shielding capabilities, and good heat transfer in forced and natural circulation, as well as neutron properties suitable for a hard spectrum. These allow for designing systems with high simplification and wide safety margins. On the other hand, opacity, corrosion, and innovative design choices require demonstration of the technology in experimental campaigns before using it in the industrial field. In recent years, Westinghouse Electric Company has begun the conceptual study of an intermediate size LFR [~460 MW(electric)] which, by exploiting the opportunities of the technology, aims at marketability over the next decade. One of the features is a passive heat removal system that allows, through different heat exchange mechanisms including conduction, convection, and radiation, for the transfer of decay heat from the reactor block to a pool of water inside the containment. The system is designed for indefinite heat removal thanks to channels that allow outside air to replace water following complete boiling. For the geometry of the system, the size, the materials, and the heat transfer mechanisms, an experimental activity is required to validate the prediction of the calculation codes and potentially support design optimization. The UK Department for Business, Energy & Industrial Strategy has recently subsidized the design, procurement, installation, and operation of the Passive Heat Removal Facility, an experimental facility to study the innovative safety system. This paper presents the activities of scaling, design, pretesting, and installation of that facility.

Three-dimensional numerical research on coupling heat transfer characteristics of PRHR HX under different water levels based on the experimental model

The passive residual heat removal system (PRHRs) is an important heat transfer system to ensure the removal of residual heat in non-LOCA conditions occurring in advanced Generation III nuclear reactors. The water level of the In-containment Refueling Water Storage Tank (IRWST) decreases due to long-term water evaporation under accident conditions and the later stages of PRHRs operation. Additionally, the water level also decreases during the water leakage accidents. The reduction of water level contributes to a decrease in the heat transfer area and the passive residual heat removal ability of the core. In order to investigate the coupling heat transfer characteristics of both primary and secondary side in the heat exchanger under different water levels, Three-dimensional numerical simulations of the PRHR HX's operation under different water levels was carried out based on the scaled-down IRWST&PRHR HX experimental model. The fluid flow and heat transfer processes inside and outside the C-type tubes were analyzed in detail. The results indicate that when heat transfer area is reduced by 50%, the overall heat transfer rate is reduced by about 41%, meanwhile the average heat transfer coefficient is about 37.4% higher than that under the full water level condition. It indicates that PRHR HX can maintain high passive residual heat removal capacity even under half water level, which provides an important reference for the design and safe operation of the PRHR HX.