Direct Reactor Auxiliary Cooling System Research Articles

A novel nonlinear radiative heat exchanger for molten-salt applications

One of the main difficulties associated with using molten salts as heat transfer fluids (HTFs) in practical applications such as in the solar and nuclear energy industries is the possibility of solidifying molten salts. Since the salts proposed for Generation-IV nuclear reactors (MSRs) have higher melting points, this is especially true for those reactors. The consequences of accidental freezing within the direct reactor auxiliary cooling system (DRACS) of MSRs could be catastrophic. DRACS is a passive safety system designed to remove decay heat from the reactor core during an emergency. The present study examines, as a case study, the freezing problem in the most vulnerable component of DRACS, namely the natural draft heat exchanger (NDHX). For this application, a unique and innovative heat exchanger design termed radiative heat exchanger (RHX) is proposed, with nonlinear heat transfer characteristics. Specifically, by suppressing convection and exploiting thermal radiation as the primary heat transfer mechanism, the RHX concept reduces the heat rejection/removal rate passively to allow safe operation when the molten salt coolant temperature drops and is near its melting point, while maintaining a high capacity for decay heat extraction in case the coolant temperature rises during an accident and the reactor becomes overheated; this moderation of the heat removal rate is accomplished without the requirement for external (manual or automatic) control. A thermohydraulic model is developed and applied to evaluate the proposed RHX with the existing conventional NDHX. Compared to the conventional design, the RHX has the potential to reduce heat removal by approximately 30 % at low temperatures and near-solidification conditions. In the case of the prototypical 20 MW reactor with a decay rate of 0.2 MW, RHX extends the freezing time of molten salt by three times, to approximately 18 h. During high-temperature events, such as the initial stages of a reactor scram, the RHX can improve decay heat extraction by 17 %. Other applications of molten salts with similar challenges can also be addressed using the proposed design approach.

Research on decay heat removal effectiveness in pool-type fast reactor CEFR based on system code SAC-DRACS coupled with CFD software

In station blackout(SBO) accident, the establishment and maintenance of natural circulation in the China Experimental Fast Reactor(CEFR) mainly relies on the decay heat removal system(DHRS). However, due to the complex structure of the reactor, it’s very difficult to model the full-scale CEFR. In the present work, DHRS and the primary circuit system of the CEFR are modeled by the System Analysis Code for Direct Reactor Auxiliary Cooling System (SAC-DRACS) code and three-dimensional CFD software FLUENT, respectively. Then the long-term transient simulation of 4000 s is calculated by coupling SAC-DRACS and CFD codes. The overall temperature stratification and local core temperature variations in the pool caused by DHRS are analyzed. Meanwhile, the residual heat removal capacity of DHRS is evaluated, the average core outlet temperature does not exceed 500.3 °C under the action of DHRS. This work provides an extension of the research method for the pool-type fast reactor safety analysis.

Performance evaluation of DRACS system of molten salt reactors using a transient solidification model

A transient solidification model was developed and coupled to ASYST code to simulate the transient behavior of the direct reactor auxiliary cooling system (DRACS) of the fluoride-salt-cooled high-temperature reactor (FHR) during the loss of heat sink (LOHS) accident. The model was firstly validated by comparing the experimental data with the simulation results that deriving from available heat transfer and friction factor correlations. The simulation results are found to be well-fitted with the experimental data. The transient solidification model successfully predicted the occurrence of solidification and the complete blockage of the heat transfer pipes in DRACSs. The solidified layer is found to be beneficial for the decay heat removal at the early stage of solidification, however, a sudden clog is found at the end stage of the solidification. Based on the simulation results, an optimization attempt was made to extend the failure time of DRACS by reducing the length of heat transfer tubes. The model is applicable to all the reactors where solidification may occur, such as other molten salt reactors (MSRs) and lead-cooled fast reactors (LFRs).

Evaluation of a freeze-tolerant decay heat removal system redundancy for fluoride salt-cooled high-temperature reactors (FHR)

Typical FHRs use a direct reactor auxiliary cooling system (DRACS) for decay heat removal under a variety of accident scenarios where active normal shutdown cooling is unavailable. Response surface method has been used for evaluating the reliability of the passive system. In this paper, a loss of heat sink (LOHS) accident sequence is evaluated using the response surface methodology. Response surfaces on maximum core outlet temperature and DRACS freezing time has been presented for evaluating the effects of friction loss, heat transfer coefficients and partial and full blockages of the DRACS. The present study shows that for one DRACS failure situations, the system is unconditional safe. But, for two DRACSs failure situations, there is probability that peak core outlet temperature (PCOT) will not exceed the allowable temperature. During the LOHS accident, the DRACS would not freeze within 3.33 days. This study will provide an insight for investigating DRACS redundancy and functional failure.

An experiment‐based validation of a system code for prediction of passive natural circulation in sodium‐cooled fast reactor

To evaluate residual heat removal capacity of next generation China sodium-cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss-of-flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR.

Code validation of a scaled-down DRACS model in RELAP5/SCDAPSIM/MOD 4.0

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available.

Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Abstract Sodium-cooled fast reactor (SFR) is one of the most promising reactors among the six Gen-IV reactor systems due to its significant advantages in close fuel cycle, comprehensive technology foundations and operation experiences. China is designing and constructing a demonstration SFR, in which, to ensure the reactor passive safety, a direct reactor auxiliary cooling system (DRACS) with the inter-wrapper flow is proposed as the decay heat removal system. Xi'an Jiaotong University is in charge of the DRACS analysis code development. The physical models in the DRACS are extracted and the numerical models for each component are established. The Gear method and the SIMPLE method are adopted as the primary solution algorithm. The code is developed and validated by EBR-II and PHENIX benchmarks, the results of which indicate that the code can predict experimental results very well.

Conceptual design of a freeze-tolerant Direct Reactor Auxiliary Cooling System for Fluoride-salt-cooled High-temperature Reactors

One of the key potential advantages of the Fluoride-salt-cooled High-temperature Reactor (FHR) is the use of passive safety systems that ensure the safe response of the reactor during anticipated operational occurrences, design basis events, and beyond design basis events. Typically, FHRs use a Direct Reactor Auxiliary Cooling System (DRACS) to remove the decay heat under a variety of accidents when the active reactor shutdown system is unavailable. Thus, it plays a key role in preventing the overheating/overcooling of the reactor in case of the failure of other primary means of heat removal. To enhance the operational reliability of DRACS, two DRACS designs are proposed in this study. The proposed DRACS designs are composed of two natural circulation salt loops and a water tank as the heat sink. This study focused on the transient behavior of DRACS of the FHRs during Loss of Heat Sink (LOHS) accident. Calculations were performed using RELAP5-3D to evaluate the design features of each concept, based on one design concept, the University of California, Berkeley’s Mark-1 pebble-bed FHR (PB-FHR). The calculation shows that the Mk-1 PB-FHR with DRACS proposed in this paper can operate for a relatively long time (about 90 h for design A and 430 h for design B) for a LOHS transient without operator involvement.

Safety Analysis of a 300-MW(electric) Offshore Floating Nuclear Power Plant in Marine Environment

The Offshore Floating Nuclear Plant (OFNP) integrates an advanced light water reactor into a cylindrical, double-hull, floating platform. It offers a series of potential benefits in economics and safety. The 300-MW(electric) version, named OFNP-300, uses an ocean-based direct reactor auxiliary cooling system (DRACS) to remove decay heat from the core passively and indefinitely during loss of feedwater or loss of off-site power events. In the ocean, the OFNP platform may roll during storms or even statically tilt following asymmetric flooding of underwater compartments. The effects of rolling motion and static tilt on the engineered safety systems are investigated in this paper using a RELAP5-3D (version 4.3.4) model of OFNP-300. The oscillations of the platform are described as the superposition of sinusoidal motions for the six degrees of freedom, i.e., heave, roll, pitch, yaw, sway, and surge. The plant’s thermal-hydraulic responses to two postulated accidents, i.e., loss-of-coolant accident (LOCA) and station blackout (SBO), are then studied in three scenarios: (a) a design-basis 100-yr storm, (b) a bounding scenario in which the platform is assumed to pitch and roll with an amplitude of 20 deg, and (c) a bounding scenario in which the platform experiences a static tilt of 30 deg. The results of the RELAP5 analysis show that the safety margins of OFNP-300 are not challenged in the aforementioned three postulated scenarios. From a thermal-hydraulic point of view, the pitch and roll motions affect the flow in the DRACS but have no negative effect on the temperatures in the core during LOCA and SBO. Static platform tilt is tolerable up to 45 deg, beyond which the emergency core cooling system can no longer function.

Analyses of decay heat removal tests with PRACS and DRACS of a sodium loop using the NETFLOW++ code

The present paper describes the analyses of decay heat removal transients conducted at a sodium facility PLANDTL. The objective of the present analyses is to confirm whether the 1-D NETFLOW++ code can simulate the natural circulation induced in the PLANDTL when a decay heat removal system is operated. The facility has 7 simulated fuel assemblies, one primary heat transport system (HTS) with a primary reactor auxiliary cooling system (PRACS), a direct reactor auxiliary cooling system (DRACS) and an air cooler to remove decay heat, secondary HTS and its cooling system with an air cooler. Two representative tests among several natural circulation tests to confirm interactions between each loop and the PRACS or the DRACS are selected for the code validation experiments. Decay heat removal systems with the PRACS and the DRACS are discussed for the next generation fast breeder reactor in Japan instead of the intermediate reactor auxiliary cooling system (IRACS) provided in the “Monju” reactor. Good agreement was obtained between the test results using the PRACS and the DRACS and the calculated results when two important thermal-hydraulic models specific for low flow rate conditions are used together with common thermal-hydraulic models.

Transient freezing of molten salts in pipe-flow systems: Application to the direct reactor auxiliary cooling system (DRACS)

The possibility of molten-salt freezing in pipe-flow systems is a key concern for the solar-energy industry and a safety issue in the new generation of molten-salt reactors, worthy of careful consideration. This paper tackles the problem of coolant solidification in complex pipe networks by developing a transient thermohydraulic model and applying it to the ‘Direct Reactor Auxiliary Cooling System’ (DRACS), the passive-safety system proposed for the Generation-IV molten-salt reactors. The results indicate that DRACS, as currently envisioned, is prone to failure due to freezing in the air/molten-salt heat exchanger, which can occur after approximately 20minutes, leading to reactor temperatures above 900°C within 4hours. The occurrence of this scenario is related to an unstable behaviour mode of DRACS in which newly formed solid-salt deposit on the pipe walls acts to decrease the flow-rate in the secondary loop, facilitating additional solid-salt deposition. Conservative criteria are suggested to facilitate preliminary assessments of early-stage DRACS designs. The present study is, to the knowledge of the authors, the first of its kind in serving to illustrate possible safety concerns in molten-salt reactors, which are otherwise considered very safe in the literature. Furthermore, and from a broader prospective, the analytical tools developed in this study can also be applied to examine the freezing propensity of molten-salt flows in other complex piping systems where standard, finite element approaches are computationally too expensive.

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.

An experimental study on natural circulation decay heat removal system for a loop type fast reactor

ABSTRACTA fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation.

Thermal hydraulics analysis of the Advanced High Temperature Reactor

The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

Scaling analysis for the direct reactor auxiliary cooling system for FHRs

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility.

DRACS thermal performance evaluation for FHR

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 and the Natural Draft Heat Exchanger. In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, and to keep the DRACS ready for activation, if needed, during accidents.To help with the design and thermal performance evaluation of the DRACS, a computer code using MATLAB has been developed. This code is based on a one-dimensional formulation and its principle is to solve the energy balance and integral momentum equations. By discretizing the DRACS system in the axial direction, a bulk mean temperature is assumed for each mesh cell. The temperatures of all the cells, as well as the mass flow rates in the DRACS loops, are predicted by solving the governing equations that are obtained by integrating the energy conservation equation over each cell and integrating the momentum conservation equation over each of the DRACS loops. In addition, an intermediate heat transfer loop equipped with a pump has also been modeled in the code. This enables the study of flow reversal phenomenon in the DRACS primary loop, associated with the pump trip process. Experimental data from a High-Temperature DRACS Test Facility (HTDF) are not available yet to benchmark the code. A preliminary code validation is performed by using natural circulation experimental data available in the literature, which are as closely relevant as possible. The code is subsequently applied to the HTDF that is under construction at the Ohio State University.

Radioactivity evaluation of the secondary sodium in DRACS of the Japan Sodium-cooled Fast Reactor

d This paper evaluates the secondary sodium activation in Direct Heat Exchanger (DHX), the radioactivity density and the dose rate around the secondary sodium pipes in Direct Reactor Auxiliary Cooling System (DRACS) of the Japan Sodium-cooled Fast Reactor (JSFR), by improving the exactness of a shielding calculation model with the Monte Carlo methodology. From this result, the prospects to meet the shielding design requirements for radioactivity free areas are obtained.

C110 高速増殖炉崩壊熱除去自然循環運転の熱流動特性

Thermal-hydraulic characteristics of the decay heat removal system (DHRS) has been evaluated by both a water test and its computational analysis. The direct reactor auxiliary cooling system (DRACS) was studied as a representative DHRS in the LMFBR design. A natural circulation flow of the DRACS is originated from the buoyancy force balanced with the pressure loss within the active core. Hence it appears that the performance of the DRACS could be fully estimated by the thermal-hydraulic study of one fuel sub-assembly under the condition of low flow rate. The inter-wrapper flow ocurring in the gap between core sub-assemblies under the natural circulation operation might have a large capacity to cool the core. An inverse flow ocurring at the outer region of the core was due to the cooling effect of the inter-wrapper flow.

An experimental investigation on penetration of buoyancy-driven cold liquid into a vertical channel with hot, forced-flow

An experimental investigation on penetration of cold liquid into a modeled subassembly channel under mixed convection, from the hot plenum of a prototypical fast breeder reactor (FBR) design has been conducted. The penetration phenomenon occurs under certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system (DRACS) for decay heat removal and can influence the natural circulation head that determines the core flow rate and consequently the core cooling rate. In the present experiment, a simplified test section simulating the upper plenum and a subassembly channel was constructed wherein both temperature and velocity measurements were made to investigate the penetration of cold water into a vertical channel with upward flowing hot water. Upon comparing temperature and velocity data we found an overall similarity in their respective spatio-temporal distributions as expected. The data suggested correlations describing the onset of flow penetration and penetration depth into the channel, expressed as a combination of Reynolds and Grashof numbers for a given Prandtl number. Flow penetration was identified based on the temperature and velocity data. This work suggests a dimensionless grouping based on an order of magnitude analysis that satisfactorily casts the experimental data while noting the subtle differences with other investigations.

Study of Thermal-Hydraulic Characteristics during Decay Heat Removal in a Pool-Type Fast Breeder Reactor

This paper reports that the thermal-hydraulic characteristics of a pool-type fast breeder reactor during decay heat removal are studied experimentally with a one-eighth-scaled three-dimensional reactor model using water as the working fluid. The model is designed to simulate reactor operations from full power to scram transient and decay heat removal conditions as well as two types of decay heat removal systems, a direct reactor auxiliary cooling system (DRACS) and a primary reactor auxiliary cooling system (PRACS). The transient thermal-hydraulic performance of the DRACS is compared with that of the PRACS for the water test results. The results are used to evaluate the accuracy of the multidimensional analysis codes AQUA and THERVIS. The analytical results are generally in good agreement with the measurements.