Decay Heat Removal System Research Articles

Optimizing the geometry of tube-in-tube Thermosyphon heat transport device from instability Perspective: A computational study

Since the Fukushima disaster, there has been a significant surge in the adoption of infinite time passive decay heat removal systems. In this context, the thermosyphon heat transport device (THTD) has distinct advantages over natural circulation loops (NCLs). Previous studies proved that the THTD experiences flow instability characterized by recirculating loops under specific geometric and operational conditions. However, identifying the critical value for these parameters remains unresolved. Therefore, the present study deals with the geometric optimization of THTD through computational (CFD) simulations.Initially, the CFD model is validated with earlier experimental data, which resulted in a minimal deviation of −2 % to 5 %. The parameters for geometric optimization of THTD are height of the system (H), the center-line elevation difference between the source and the sink (Δz), heater length (Lh), cooler length (Lc), hot leg length (Lhl), and area ratio (AR) between hot leg and cold leg, under the heater power range of 100 W to 1000 W. The impact of these geometric and operational variations on flow instability is evaluated by observing the presence of recirculation loops, which are identified by analyzing velocity vectors and contours. The key findings of the present study are: flow instability in the form of a single recirculating loop sweeping radially along the cross-section of the annulus, thereby leading to radial flow maldistribution; a center-line elevation greater than 0.8 m eliminates the flow instabilities; interchanging heating and cooling lengths for a given center-line elevation doesn’t alter the criteria for instability; Heater power of 100 W leads to instability irrespective of center-line elevation; the optimal cooling length can be considered as 5 % to 25 % of the total height of the THTD, whereas the optimal hot leg length can be considered as 25 % to 45 % of the total height; and the HTF flow rate increases with a rise in the area ratio until AR=2, after which a flow anomaly occurs, rendering the system unstable. In addition to these aspects, this study also suggests optimal geometric features of THTD to reduce instability and attain perfect unidirectional flow. By integrating findings from current research and past studies, the design methodology for conventional THTD is comprehensively presented to the scientific community.

Design and analysis on the HP-PHRS for small modular lead-bismuth fast reactor

The heat pipe is an efficient heat-conducting element that can efficiently and passively transfer heat from a high-temperature heat source to a low-temperature heat source. It is widely used in the various industries. Based on high-temperature heat pipe technology, this paper designs a passive decay heat removal system for a small modular lead–bismuth fast reactor, and uses the system code ATHLET coupled with the heat pipe calculation module to calculate and analyze the thermal–hydraulic characteristics of the reactor under station black out accident.

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

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.

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.

Performance analysis of heat pipe-based passive in-core decay heat removal system for the small modular reactor design with MARS-KS code

This study explores the effectiveness of using Passive IN-core Cooling Systems (PINCs) for Small Modular Reactors (SMRs) during Station Blackout (SBO) conditions. An introduction to the concept of PINCs, along with strategies for their application to SMRs with design modifications, is presented. Transient safety analysis was conducted using the MARS-KS code on NuScale SMR designs that incorporate PINCs. The findings demonstrate that PINCs can serve as an additional means for decay heat removal in addition to the existing safety systems. The feasibility of PINCs as an effective means for indefinite cooling of SMRs was affirmed under SBO conditions, even in the absence of DHRS and ECCS operations. The PINCs system demonstrates its potential as an effective means for pressure reduction and long-term or indefinite cooling in SMRs under extreme conditions such as SBO.

Protected Loss-of-Flow Accident in a HLM Pool Facility: Long-Term Cooling Experiment

Since the 2000s, the development of Generation-IV fast reactors cooled by heavy liquid metals (HLMs) has been pursued by several research activities and projects, many of which are co-funded by the European Commission. One of the key points of HLMs regards their good neutronic and thermophysical properties, allowing for the of design cores with a high pitch-to-diameter ratio. In terms of passive safety, with a properly designed configuration, it is possible to increase the system capability to remove the decay power in a natural circulation regime, reducing active safety systems involvement. Such a safety-related aspect has been experimentally investigated at the ENEA Brasimone Research Center within the European Union co-funded Euratom H2020 SESAME project. An experimental campaign reproducing protected loss-of-flow accident (PLOFA) scenarios has been executed on CIRCE, a lead-bismuth eutectic-cooled pool-type facility reproducing in relevant scale the main components of HLM-cooled fast reactors. A test section named HERO has been installed in the CIRCE main vessel, hosting a double-wall bayonet tube steam generator scaled 1:1 in length with respect to the one envisioned for the ALFRED reactor. The tests reproduce the loss of primary coolant flow, while the power supplied by the heating source decreases according to a characteristic decay heat curve. The feedwater in the secondary system is regulated to operate the main steam generator as a decay heat removal (DHR) system. This paper presents the PLOFA transient reproducing the worst case, where the steam generator feedwater is suddenly stopped, simulating the full loss of the heat sink (no DHR). The main phenomena occurring during the transition from forced to natural circulation are presented and discussed. The experiment shows that, despite the loss of the forced circulation regime in the primary loop and the full loss of the heat sink, the entire system is still capable of operating safely, assuring an effective long-term cooling, as long as the thermal heat losses from the main vessel balance the decay power supplied by the heating source.

Transient analysis of SIRIO using RELAP5/MOD3.3 system code

The main outcome of the present paper is the feasibility analysis of SIRIO (Sistema di rimozione della Potenza di decadimento per Reattori InnOvativi) facility with conditions based on those of its reference facility. The aim of SIRIO project is to study an innovative Decay Heat Removal System (DHRS) for liquid metal reactor and advanced Light Water Reactor (LWR). Such system must ensure passive control of the power removed from the primary system in abnormal condition, and must ensure reactor cooling in both short and long term. This study present numerical simulations developed with RELAP5/MOD3.3, of two operational procedures: the first one is a steady-state and the second one is a transient phase with decay heat generation. The thermal-hydraulic model, developed with RELAP5/MOD3.3, simulates the whole facility including lines, valves, water and gas tanks, and the Molten Salts (MS) gap. Since there is not experimental data, the present paper is a pre-test study based on SIRO facility design.

Unique safety features and licensing requirements of the NuScale small modular reactor

Small modular reactors (SMR) offer a novel approach to the construction and operation of nuclear power plants. The NuScale VOYGR™ plant uses a simplified SMR design that is based on proven light-water reactor technology with substantial improvements in nuclear safety. It consists of a 250 MWt reactor core housed with other primary system components in an integral reactor pressure vessel surrounded by a steel containment vessel, all of which is immersed in a large pool of water that also serves as the ultimate heat sink. At the core of the NuScale safety case are three primary safety systems: the decay heat removal system, the emergency core cooling system, and the containment. The ability of the NuScale Power Module (NPM) passive safety systems to remove core decay heat for an unlimited duration is demonstrated through analysis of a beyond-design-basis extended loss of AC power with no replenishment of water to the ultimate heat sink or operator actions. The NuScale methodology to evaluate an indefinite loss of AC power uses the proprietary NRELAP5 systems analysis computer code. Analysis results show that the reactor coolant system liquid level above the core is maintained and that containment pressure remains below the vessel design pressure. Once full passive air cooling is established, containment pressure and temperature will decrease over time with decreasing core decay heat. NuScale received standard design approval in September 2020 and design certification in January 2023 for its 50 MWe NPM configured as a 12 module plant. NuScale is currently seeking standard design approval to increase its core power to 250 MWt, nominally 77 MWe per module, in a 6-module plant (VOYGR™-6) configuration. The high-level safety of NuScale’s SMR technology is foundational to a new standard of nuclear power plant resilience.

Design and Assessment of the New CIRCE-THETIS Facility for the Development of Heavy Liquid Metal Cooled Fast Reactors

Within the roadmap for the technological development of Generation IV reactors, the HORIZON2020 European Union–funded Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action (PATRICIA) project was launched to support innovative solutions for the development of the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) accelerator-driven system concept and lead fast reactor technologies in general. The ENEA contributes to the project by involving the experimental infrastructures of the Brasimone Research Center (Italy). In particular, a large-scale pool-type facility named CIRColazione Eutettico (CIRCE), using lead-bismuth eutectic as the primary coolant and pressurized water as the secondary fluid, is under refurbishment with the implementation of a novel test section (TS) named Thermal-hydraulic HElical Tubes Innovative System (THETIS) to be installed in the CIRCE main vessel. The new TS will include a vertical mechanical pump for primary coolant circulation and a new prototypical helical coil steam generator (HCSG). This steam generator concept turns out to be very promising for nuclear power plants since the helical geometry is very compact and it assures high power removed, taking up a minimum amount of space. Accordingly, with the aims of the project, the experimental tests in CIRCE-THETIS will focus on (1) investigating the thermal-hydraulic behavior of the system in steady-state operation (forced circulation regime) during operational and accidental transients (postulated scenarios) and in a natural circulation regime considering as heat sink the HCSG (acting as a decay heat removal system) and the reactor vessel auxiliary cooling system in stand-alone or coupled operation and (2) characterizing the performance of the HCSG. The present work presents the layout of the CIRCE-THETIS facility at the end of the final design phase, describing in detail the main components of the TS, along with the instrumentation installed. Focus will be given to the HCSG mock-up, for which pretest analyses using the system thermal-hydraulic code RELAP5/Mod3.3 and a computational fluid dynamics code have been carried out to support the design of the component and to evaluate its thermal-hydraulic performance under the operative conditions foreseen during the experiments.

Thermal-hydraulic optimization of a proposed EU-DEMO hydrogen passive removal system

As the R&D of magnetic fusion power demonstrating plants are approaching important steps toward concept designs, analysts are working parallelly on the safety assessment of such concepts to identify any potential risk. One of the safety concerns involves the confinement of radioactive substances during normal operation and accidental conditions. Several accident sequences inside the tokamak vacuum vessel or pressure suppression systems are characterized by the risk of hydrogen buildup and subsequent ignition that could threaten the structural confinement integrity. In the Safety and Environment work package of the EUROfusion consortium, possible approaches to mitigate the hydrogen explosion risk are under investigation. One of the exploratory solutions is based on limiting the hydrogen concentration that could reach flammable gas mixture conditions and using Passive Autocatalytic Recombiners (PARs) installed into the atmosphere of the pressure suppression systems tanks to recombine hydrogen.This paper examines the theoretical effectiveness of the PARs intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled Lithium Lead (WCLL) concept of EU-DEMO, using an optimization methodology. The involved systems have been modelled in MELCOR to estimate the PARs recombination capability as a function of the thermal-hydraulic parameters of the suppression tanks. Furthermore, the optimizer entity of the RAVEN tool is applied to perform optimization studies on the hydrogen recombination system design parameters. The goal is to explore the geometrical and thermal-hydraulic parameters that maximize the capability of the hydrogen removal system for the WCLL concept.

Simulation of decay heat removal by active means following emergency shutdown in SFRs

Decay heat removal (DHR) is an important safety function in sodium-cooled fast reactors (SFRs). Dedicated systems are provided in the plant for performing DHR functionalities. Transients following reactor shutdown need to be studied to understand the dynamics of DHR systems, design the systems, and establish reactor safety. DHR systems can operate by active as well as passive means, thanks to the phenomenon of natural circulation. In the present work, a DHR system riding on the Balance of Plant (BoP) of a typical medium-sized SFR has been modeled and incorporated into the in-house developed plant dynamics code DYANA-P. DYANA-P already has models of primary and secondary circuits of SFR. With this code enhancement, the simulation of DHR in an integrated manner became possible. In particular, thermal models of the Moisture Separator Tank (MST), Decay Heat Removal Condenser (DHRC), and hydraulic model of the DHRC flow circuit were developed. Two-phase one-dimensional mass, momentum, and energy conservation equations with homogeneous flow assumptions were used to create the models. Simulation of SCRAM in a typical SFR has been carried out to demonstrate the applications of the modified code. Two different DHR deployment schemes were studied and compared – 1. Using DHR systems designed for high pressure of ∼170 bar, 2. Using DHR systems designed for low pressure of ∼15 bar. In the first scheme, reactor cooling was achieved at the desired rate of ∼20 K/h. In the second scheme, the reactor cooling rate will be higher at ∼60 K/h.

Passive Hydrogen Recombination during a Beyond Design Basis Accident in a Fusion DEMO Plant

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

A Fan-Controlled Sodium-to-Air Heat Exchanger Configuration for Failsafe Decay Heat Removal in Sodium-Cooled Fast Reactors

Passive safety systems help to improve overall plant safety, reliability, and resilience. However, the real gain in the use of passive systems depends on the robustness of the design utilizing the passive process and the role of active elements, if any. In this paper, we propose a fan-controlled sodium-to-air heat exchanger (AHX) system design for a failsafe and passive decay heat removal (DHR) function in a pool-type sodium-cooled fast reactor. The proposed system uses a fan to control air flow and minimize heat loss during normal operation, and when the fan trips due to loss of power or a trip signal, DHR gets enabled in a failsafe mode. The system is analyzed with the help of a simplified one-dimensional model as well as with detailed computational fluid dynamics software. It is found from analysis that it is possible to control and maintain the air flow to about 4% to 5% of full flow, as in the case of conventional dampers, to minimize heat loss during normal reactor operation. The reliability of the proposed system is also analyzed and shows that the fan-controlled AHX-based decay heat removal system (DHRS) has a much better reliability compared to the conventional passive DHRS with active damper-dependent operation.

OECD/NEA/CSNI/WGAMA PERSEO benchmark: Main outcomes and conclusions

In the framework of the OECD/NEA/CSNI/WGAMA, an activity on the “Status report on thermal–hydraulic passive systems design and safety assessment” has been conducted. Within this activity, a benchmark exercise, based on the experimental data developed in the full scale PERSEO (in-Pool Energy Removal System for Emergency Operation) component separate effect test facility, built at SIET (Piacenza, Italy), has been proposed and carried out. An “OPEN” benchmark exercise, hosted by ENEA, has been conducted. Twelve results from eleven Organizations were submitted. PERSEO is a full-scale separate effect test facility designed to study a new passive decay heat removal system operating in natural circulation. Test 7 is a full pressure test (7 MPa) and investigates the system stability and the system operation. The accuracy of the calculated results has been evaluated both qualitatively and quantitatively. The latter has been conducted adopting the Fast Fourier Transform Based Method. The present paper summarizes the main features of the PERSEO facility and Test 7 and discusses the main results and outcome of the benchmark exercise.

Numerical simulation of natural circulation characteristics under different DHX layout schemes in pool-type SFR during station blackout accident

Decay heat removal system (DHRS) is an essential system for pool-type sodium-cooled fast reactor (SFR) to ensure natural circulation and remove decay heat during station blackout (SBO) accident. Decay heat exchanger (DHX) is the key equipment in DHRS. In China Experimental Fast Reactor (CEFR), DHX layout scheme consisting of two DHXs in hot pool (DHX Layout 1) is adopted. Different DHX layout scheme is designed in the advanced pool-type SFR, for example DHXs are installed in both cold and hot pool (DHX Layout 2). It is necessary to analyze the influence of different DHX layout schemes. In this paper, two DHX layout schemes were proposed. Numerical simulation was carried out, of which method was preliminarily validated. The numerical results show that DHX Layout 2 can reach long-term cooling stage of natural convection earlier than DHX Layout 1 with strengthened flow rate, resulting in more uniform temperature distribution in the hot pool. There is a local asymmetrical temperature distribution in the cold pool for DHX Layout 2. These findings by numerical simulations provide important numerical reference for the subsequent DHRS design of the advanced pool-type SFR.

Advanced Safety Programmable Logic Controller and Operational Programmable Logic Controller for Data Acquisition Control System of SADHANA Sodium Loop

The SAfety Decay Heat Analysis in NAtrium (SADHANA) sodium loop is designed with two PLC controllers. All safety control interlocks are monitored by a Safety programmable logic controller (SPLC), and temperature control, including averaging ON/OFF heaters, is managed by an Operational PLC (OPLC).The decay heat exchange system of the SADHANA sodium loop is evaluated using these combined safety and operational PLCs. In the case of reactor loss of offsite and onsite power, decay heat generated in the reactor core after shutdown is removed by four Safety Grade Decay Heat Removal (SGDHR) systems [1,2,3,4,5]. Each exchanger (AHX) is located outside the reactor containment building, and a sodium loop circulation is created by natural convection [6,7,8,9,10]. The purpose of construction of SADHANA loop, with simulated models of DHX and AHX to study the performance of the SGDHR system, has been planned [11,12,13,14,15]. The SADHANA sodium loop has been commissioned, operating at isothermal temperatures of 300°C and hot temperatures of 500°C/& 550°C. The smart instrumentation and control system comprising Loop Operational Data Acquisition and Control System (LODACS) and Experimental Data Acquisition System (EDAS) are used in the SADHANA sodium loop. During the tests, LODACS personal computer and Instrumentation panel monitor loop parameters such as sodium level, leak, and pressure. In SADHANA loop DHX and AHX temperature decay heat changes monitoring by EDAS. The temperature of the pool in Test vessel 4 (TV-4) is controlled by immersion type sodium heaters via PID controllers. Currently, the loop consists of 64 surface heaters, 30 immersion type sodium heaters, and 205 thermocouples (84 Nos. of operational thermocouples and 121 Nos. of experimental thermocouples). According to the experimental thermocouples are being monitored by a separate fast data acquisition (scan rate of 10 seconds) by in-house developed software. This technical paper also describes the implementation of PLCs in different data acquisition control systems.

Scaled validation test for high prandtl number fluid mixed convection between parallel plates

The Kairos Power fluoride salt-cooled high temperature reactor (KP-FHR) passive decay heat removal system performance strongly depends on heat transfer characteristics in the reactor vessel’s down comer region. This region is an annular gap where the salt coolant flows downward between the core barrel and reactor vessel before entering the core region at the bottom of the vessel. Kairos Power’s principal safety analysis code, KP-SAM, extrapolates from existing heat transfer correlations for this geometry and applicable flow configurations. Main flow regimes of interest for safety analysis of the KP-FHR, when the system transitions from forced to natural circulation in the vessel upon loss of forced cooling, are the laminar and transition regimes. These regimes are not very well characterized in the literature, so in order to validate the inclusion of appropriate correlations in KP-SAM, scaled separate effects tests are conducted using a test facility that matches the Reynolds, Rayleigh and Prandtl numbers between the prototypical reactor coolant, Flibe, and a surrogate fluid, Therminol VP-1. The specific boundary conditions investigated are both symmetric heating and asymmetric heating. The data generated from this test serves to develop a validated heat transfer correlation for the laminar and transition regimes for implementation in KP-SAM, in support of safety analysis and licensing of the KP-FHR.

Neutronics and thermal-hydraulics coupling analyses on transient and accident behaviors of molten chloride salt fast reactor

ABSTRACT Neutronics and thermal-hydraulics coupling transient analyses have been carried out to investigate the intrinsic safety characteristics and fuel temperature ranges of a molten chloride salt fast reactor. The analysis system includes a primary heat transport system and a secondary heat transport system connected by aheat exchanger and a decay heat removal system connected to the secondary system. The neutronics and thermal-hydraulics coupling analysis was performed with the RELAP5-3D code to analyze the entire plant behavior of each event, the temperature and velocity of the reactor inlet flow were obtained, and the reactor power was analyzed in detail following the FLUENT code. Through these analyses, the most severe events for the reactor were revealed. Analysis using the two codes showed that the reactor power calculated by RELAP5-3D was in good agreement with that of FLUENT although the detailed flow inside the core could not be reproduced.