Passive Residual Heat Removal System Research Articles

Analysis of transient characteristics and design improvement of the passive residual heat removal system of NHR-200-II

NHR-200-II is a small integrated pressurized water reactor with 200 MW core thermal power. The core heat is transferred to two independent intermediate circuits via fourteen in-vessel primary heat exchangers (PHE), and the heat in the intermediate circuits is transferred to feedwater by two steam generators (SG) in the two intermediate circuits respectively. A passive residual heat removal (PRHR) branch is connected to each intermediate circuit to remove core decay heat under postulated accidents. During normal operation, PRHR branches are isolated by valves while SG branches in intermediate circuits are open. The valves in PRHR branches will be opened and the isolation valves of SG branches will be closed during decay heat removal scenarios. The decay heat removal capacity of NHR-200-II PRHRS could be seriously deteriorated once the isolation valves for SG branches fail to close, which was confirmed in a scaled integral test loop previously. Current understanding of PRHRS’s thermal-hydraulic characteristics with possible isolation failure in SG branches is limited. In this paper, the NHR-200-II PRHRS is modeled with RELAP5 considering the case of success and fail to isolate SG branches. A series of numerical simulations are carried out to study the impact of various parameters, such as the initial temperature, the size of the intermediate circuits’ header, and the initial flow direction in the intermediate circuits. Oscillatory flow is found when SG branches fail to be isolated under certain parameters combinations. An improved PRHRS design is purposed to eliminate possible flow oscillations, and the purposed improved design are tested by numerical simulations.

Experimental investigation on steam condensation characteristics in horizontal heat transfer pipe of PRS for HPR1000

The heat exchanger is an important component of the passive residual heat removal system (PRS) in the Hualong Pressurized Reactor (HPR1000). The heat released via steam condensation inside the heat transfer pipes is delivered to water stored outside of the containment. The heat transfer along the heat transfer pipes is systematically studied experimentally. Three horizontal test sections with different dimensions are designed and employed in the experiment. The test data are obtained under different pressure, steam quality, and mass flow rate conditions and compared with existing steam condensation correlation. The test results reveal that the heat transfer coefficient decreases from 41.5 kW/(m2·K) to 24.0 kW/(m2·K) with the pressure increased from 1.0 MPa to 8.3 MPa. In the higher pressure conditions, the larger condensed flow rate provides larger thermal resistance, decreasing the condensation heat transfer coefficient. The steam condensation heat transfer coefficient is increased from 14.3 kW/(m2·K) to 41.4 kW/(m2·K) as the steam quality is increased from 0.07 to 0.70. The condensation heat transfer coefficient is increased from 28.6 kW/(m2·K) to 45.8 kW/(m2·K) as the mass flow rate is increased from 0.029 kg/s to 0.072 kg/s. The maximum prediction deviation of Shah’s correlation for all test data is ±34.9 %.

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.

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.

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.

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.

Design and transient analysis of a compact and long-term-operable passive residual heat removal system

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 °C/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

Numerical study on melting process and heat transfer analysis of a multitube combination latent heat storage unit with innovative harpoon-type fins

The heat transfer rate of a coupled system combining a passive residual heat removal system and a latent heat thermal energy storage system is directly related to worker safety in nuclear power plants. To enhance the heat transfer performance of the coupled system, in this study, the melting process of phase change material in the latent heat thermal energy storage system is numerically investigated by inserting innovative harpoon-type fins. A new assessment index is also proposed to evaluate the comprehensive heat transfer performance of fins. More than thirty cases with different harpoon-type fins are designed, and the impacts of various efficient structural parameters of the harpoon-type fin on the melting process are evaluate. The response surface method is used to optimize the structure. The results demonstrated that the handle length and the number of fins significantly improve the heat transfer performance of the harpoon-type fin. The optimal geometrical parameters (arc angle θ = 45°, handle length Lha = 10 mm, fin number Nf = 8, and arc number Na = 1) of harpoon-type fins are obtained. The heat transfer rate of the optimal fin structure is six times that without fins, and the total melting time is shortened by 85.9%. Furthermore, the longitudinal extension of the fins is more important than the complex horizontal extension of fins. The harpoon-type fin not only shows better performance in the coupled system but can also be used in shell-and-tube latent heat thermal energy storage systems.

Thermal Hydraulic Analysis of a Novel Passive Residual Heat Removal System with Loop Thermosyphons for the Pool-vessel Reactor

Advanced pool-vessel reactors require reliable, secure, and passive mechanisms to discharge residual heat from the reactor following shutdown. For this reason, this paper proposes a novel Passive Residual Heat Removal System (PRHRS) with loop thermosyphons. To evaluate the residual heat removal and heat transfer capabilities of the PRHRS, a numerical analysis tool is developed. To achieve this, A thermal hydraulic model is established by secondary development of ANSYS FLUENT, designed to couple the primary loop, reactor pool, and loop thermosyphons. A feasibility study is conducted utilizing the proposed model, and the results demonstrate that this system can efficiently and promptly transfer residual heat to the reactor pool and ultimately to the atmosphere, preventing the core from overheating as well as the pool water from saturated boiling. This model is highly effective in analyzing the thermal hydraulic and safety characteristics of pool-vessel reactors and providing technical support for the design and optimization of the PRHRS.

Experimental study of steam condensation heat transfer characteristics in a horizontal tube under rolling motion

As an efficient heat transfer method, in-tube condensation heat transfer is widely used in various fields, including passive residual heat removal systems (PRHRS) of floating power plants under ocean conditions. The experimental investigation aimed to study the condensation heat transfer characteristics of steam in a 25 mm inner diameter horizontal tube with a heat transfer length of 1050 mm under rolling motion. The study used a mass flux of 30 kg/(m2·s) and investigated average vapor qualities, rolling amplitude, and rolling period ranging from 0.32 to 0.79, 0 °–20°, and 10 s–20 s, respectively. The experimental results show that the condensation flow patterns under rolling conditions are more complex than those under non-rolling motion due to the influence of the rolling period and rolling amplitude. The instantaneous heat transfer coefficient (IHTC) changes periodically due to the rolling motion and has the same change period. The pulsation amplitude of the IHTC increases with increasing rolling amplitude but weakens as the vapor quality increases. The effect of the rolling period on the IHTC is much less than that of the rolling amplitude. At low vapor quality, the time-averaged heat transfer coefficient (THTC) under the rolling motion condition increases with the rolling amplitude and period, and the maximum enhancement of THTC is 44%. As the vapor quality increases, the THTC under rolling conditions is the same as that under non-rolling conditions. These findings provide important insights into the condensation heat transfer characteristics of vapor in a horizontal tube under rolling motion, which can help improve the design and operation of in-tube condensation heat transfer systems in various fields.

Experimental study on the influence of inclination on operating characteristics of secondary side passive residual heat removal system

The present work discusses the operating characteristics of the secondary side passive residual heat removal system (PRHRS) of an offshore floating nuclear plant (OFNP). A scaled experimental facility was built on a platform that can be inclined to a certain angle. Experiments were conducted in horizontal and various inclined conditions, and the transient response characteristics and steady-state operating characteristics of the PRHRS were obtained. The experimental results reveal that different inclined directions will change the potential height difference between the cold and heat sources and affect the natural circulation capacity of the system. In general, due to the small inclination of the floating platform in the ocean, the residual heat can be effectively exported by PRHRS. Thermal stratification in the cooling water tank has significant effect on the evolution process of outlet temperature of the passive residual heat removal heat exchanger, and the influence of thermal stratification on the outlet temperature of the heat exchanger decreases under inclined conditions. The research results in this paper can provide some guidance for the design of operating rules of the secondary side PRHRS.

Simulation and design of passive residual heat removal system for Fluoride-Salt-cooled high-Temperature Advanced reactor (FuSTAR)

The new generation of reactor technology: Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), mainly applied to the steady supply of stable electricity and high-temperature process heat for remote, underground, and arid areas. So far, the design of neutron physics and thermal hydraulics of the core have been finished, but the heat transport system and Passive Residual Heat Removal System of FuSTAR still need to be designed and optimized. Given the complexity and consistency of the design, a unified design and optimization method is required, which can directly obtain detailed configurations and dimensions. In this paper, a design and optimization method - Multi-Layer Nonlinear Programming was proposed to obtain the optimal parameters, including vital thermodynamic values, the dimensions of heat exchangers, the configurations of pipelines, and the optimal equivalent parameters in the transient simulation. Finally, based on Multi-Layer Nonlinear Programming, the effectiveness of the Passive Residual Heat Removal System was demonstrated by the deterministic safety analysis. The detailed parameters of the Passive Residual Heat Removal System in this paper can provide references for subsequent experimental verification. Moreover, for any reactor relying on a heat transport system, researchers can directly obtain its excellent preliminary configurations by this method avoiding a lot of iteration and conflict design, which is conducive to the integrated modeling and simulation of a new generation of reactors.

Experimental investigation on characteristics of Multiple-Loops coupled natural circulation system

The passive safety design is advantageous because it simplifies the system and brings higher safety features. Passive residual heat removal system (PRHRS) is one of the most important passive safety systems, which is designed to cool down the reactor during accidental conditions. For floating nuclear power platforms, seawater could be utilized as the final heat sink of PRHRS through multiple-loops coupled natural circulation.To study the operating characteristics and thermal hydraulic behavior of multiple-loops coupled natural circulation system, the Natural Circulation Mechanism Experimental System (NCMES) has been constructed. The station blackout accident (SBO) scenario is simulated in NCMES, which leads to the stopping of circulation pump and operation of PRHRS. Based on the experimental data, an analysis was carried out using the thermal hydraulic system code RELAP5, aiming to validate applicability and accuracy, as well as predict the operating characteristics of the experiment facility. The comprehensive analysis of experimental and calculated results shows that when it comes to the SBO scenario, the releasing residual heat and sensible heat of fluid can be removed to seawater through triple-loops coupled natural circulation; the calculated results fit the experimental data quite well and the RELAP5 code is applicable to predict operating characteristics of multiple-loops coupled natural circulation and PRHRS. As a result of self-adaptive characteristic of natural circulation, with the heating power decreasing, the response rate of natural circulation flow rate and fluid temperature is slow.

Experimental and analytical investigation on SMART passive safety systems for three 2-inch SBLOCA tests using SMART-ITL

The SMART-ITL facility is an integral test loop for SMART, a 360 MWt integral type reactor. Using this facility, several design-basis and performance tests were conducted to validate the passive safety systems of SMART. Among these, three 2-inch small-break loss-of-coolant accident tests were introduced, where the operating conditions of the passive safety injection system (PSIS) and passive residual heat removal system (PRHRS) were different. The main difference between the test results was in the reactor vessel (RV) level. Both PSIS and PRHRS contributed to the RV levels. The tests were also simulated using a system analysis code, MARS-KS. Using the code, major parameters were predicted reasonably well. The physical phenomena behind core makeup tank (CMT) injection were explored in depth by referring to the code calculation results, and the suspension of CMT injection was simulated by considering the heat loss at the CMT wall.

Experimental research and theoretical analysis of flow instability in advanced secondary passive residual heat removal system

To cope with station blackout (SBO) accident in the pressurized water reactor (PWR) nuclear power plant (NPP), the advanced secondary passive residual heat removal system (ASP) was developed. A newly advanced secondary passive heat removal system test facility (ASPTF) for generation III NPP was designed and constructed to obtain the thermal hydraulic characteristics of ASP. The flow instability in ASP was observed in ASPTF, which was more likely occurred under low heating power and system pressure conditions. The effects of different influencing factors on flow instability in ASP were obtained. Then, the theoretical analysis of flow instability in ASP was conducted. Based on the experimental and theoretical analysis results, the optimization and improvement measures of ASP were summarized.

Reliability assessment of passive residual heat removal system using FORM methodology with chaos control

Passive safety systems are adopted extensively to increase the inherent safety of the nuclear reactor. However, their stability of operation is more susceptive to thermal hydraulics uncertainty due to the relatively small driving force of passive systems. Coupled with a shortage of pertinent experimental and operational data, its reliability assessment is still a crucial and challenging task. So far, there are no accepted techniques for this problem and they are not yet frozen. Hasofer-Lind and Rackwitz-Flessler method (HL-RF) is an efficient iterative approach to identify the most probable point (MPP) and calculate the reliability. However, the approach may suffer from instability when dealing with highly nonlinear limit state functions (LSFs) or high-dimensional problems. In this paper, a straightforward and efficient method depending on chaos control was proposed to deal with numerical instability and chaos non-convergence problems. The proposed method was utilized to assess the reliability of the passive residual heat removal system (PRHRS) in Integral-type Pressurized Water Reactor (IPWR200). The results showed that it could efficiently and reliably locate MPP and accurately obtain the reliability index.

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Nuclear energy plays an important role in the green low-carbon energy. The lead-based reactor is one of the most popular small modular reactor types. The pre-conceptual design of a 100 MW(e) small modular lead-bismuth eutectic cooled reactor was developed by Shanghai Nuclear Engineering Research & Design Institute CO., LTD. to meet the market demand of advanced small-scale nuclear plant. Representative accidents were analyzed for the 100 MW(e) small modular lead-bismuth reactor using ATHLET/MOD 3. The result shows that under the design basis conditions, the performance of the reactor can meet the safety criteria with the protection system working normally. The coolant solidification in the primary circuit under these transients should be prevented. Under the design extension condition, the highest cladding temperature is much lower than the melting point. In the unprotected reactivity insertion condition, there may be a short time of local melting of the fuel, which also meet the acceptance criteria. The safety analysis demonstrated the passive safety of the design with the negative temperature coefficient, strong natural circulation capability, and the passive residual heat removal system.

Analysis of Influence of Tube Failure on Performance of Passive Residual Heat Removal Heat Exchanger

A passive residual heat removal system plays an important role in cooling the reactor core under accident conditions. The computational fluid dynamics (CFD) software package ANSYS Fluent is used to analyze the influence of malfunction of any 2 of 12 tubes of a passive residual heat removal heat exchanger (PRHR HX) on its performance. Then the computation was validated using the published experimental data. Five different scenarios and a normal condition are computed to analyze the influence of locations of the malfunctioning tubes on the heat transfer performance of the PRHR HX. The results show that the tube defect reduces the amount of heat transferred by the PRHR HX. However, it is correlated with the size of the surface area of the deficient tubes instead of their locations. In other words, analysis suggests that defect tubes with the same surface area should result in similar damage regardless of the location.

Experimental Study on Singlephase Natural Circulation Flow and Heat Transfer of Passive Residual Heat Removal System of Swimming Pool-type Low-temperature Heating Reactor

Experimental Study on Singlephase Natural Circulation Flow and Heat Transfer of Passive Residual Heat Removal System of Swimming Pool-type Low-temperature Heating Reactor