Passive Decay Heat Removal Research Articles

Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the containment outside into integrated passive safety tanks (IPSTs). The coolant in the IPSTs is used for various functions in accident scenarios. Also, potential problems for the realistic installation of the IPSS are proposed and the solutions to these problems are schematically described. IPSS is the design for the passive safety enhancement in preparation for a loss of AC power. Consequently, it is designed for the supplementation and enhancement of current nuclear power plants, not as a replacement. The specific optimization design for each current or future reactor will be studied as further works.

Reduction in thermal stratification in two phase natural convection in rectangular tanks: CFD simulations and PIV measurements

Heat transfer by natural convection in a cavity is encountered in various industrial applications, such as heating and cooling of living spaces, fire in building, solar thermal collector systems, electronic and photovoltaic cooling devices, thermosiphon heat exchangers, passive decay heat removal systems, air condensers, etc. In such systems, the cooler (heavier) fluid naturally moves towards the bottom of the tank while the hotter lighter fluid rises towards the top. These flows result into certain extent of non-uniformity (or stratification) of temperature/concentration depending upon the intensity of flow. We have carried out flow (using particle image velocimetry) and temperature (using thermocouples) measurements in a rectangular tank (0.8×0.6×0.6m3) fitted with a central tube (forming the heat transfer surface). In order to reduce the stratification, the effects of various internals have been examined. These include (a) changing the ratio of area (heat transfer) to volume (aspect ratio), which is achieved by insulating the heating tube with the help of teflon, (b) introduction of draft tube concentric to the heat transfer tube, and (c) provision of non-conducting or highly conducting fins attached to the tube at different axial locations. Additionally, computational fluid dynamic (CFD) simulations of these systems were performed using the commercial software FLUENT-6.3. The extent of stratification has been investigated for a wide range of Rayleigh numbers (4.34×1011≤Ra≤2.59×1014). In addition, the flow information obtained from PIV was analyzed for insights into the dynamics of turbulent flow structures. For this purpose, we have used the signal processing technique of Discrete Wavelet Transform (DWT). From the analysis, we were able to estimate the size, velocity and energy distribution of turbulent structures. This detailed knowledge was employed in surface renewal type of theories for the estimation of rates of heat transfer. A good agreement was observed between the predicted and the experimental values of heat transfer coefficient.

A probabilistic model for online scenario labeling in dynamic event tree generation

Dynamic event trees provide a wide coverage of possible system evolution sequences (scenarios) and may lead to the simulation of thousands of scenarios following a single initiating event. The large number of scenarios can be a burden in terms of computational time and storage requirements. However, not all of the scenarios are equally significant. From a safety point, failure scenarios or the scenarios that lead to undesirable consequences are more important than the scenarios that represent normal system evolution (non-failure scenarios). A method is presented for online labeling of non-failure scenarios. Since the number of non-failure scenarios is usually much larger than that of failure scenarios, substantial computational savings could be obtained if the non-failure scenarios can be identified and not pursued by the simulator. First, the parameters of a hidden Markov model that represents the behavior of non-failure scenarios are learned using the examples of the non-failure scenarios. Next, the failure behavior with respect to the non-failure model is learned using sample failure scenarios. During the succeeding system simulations, a scenario is labeled as non-failure if its evolution path is more likely to fit the constructed model than the learned failure behavior. Experiments using RELAP5/3D model of a fast reactor utilizing an RVACS (Reactor Vessel Auxiliary Cooling System) passive decay heat removal system show that the proposed method is capable of correctly labeling over 85% of non-failure scenarios without mislabeling the failure scenarios and provide time savings of at least 55%. We also investigate the sensitivity of the proposed labeling scheme depending on the number of hidden states in HMM and the nature of the state variables used for scenario representation.

Counterpart experimental study of ISP-42 PANDA tests on PUMA facility

A counterpart test to the Passive Nachwärmeabfuhr und Druckabbau Test Anlage (Passive Decay Heat Removal and Depressurization Test Facility, PANDA) International Standard Problem (ISP)-42 test was conducted at the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. Aimed to support code validation on a range of light water reactor (LWR) containment issues, the ISP-42 test consists of six sequential phases (Phases A–F) with separately defined initial and boundary conditions, addressing different stages of anticipated accident scenario and system responses. The counterpart test was performed from Phases A to D, which are within the scope of the normal integral tests performed on the PUMA facility. A scaling methodology was developed by using the PANDA facility as prototype and PUMA facility as test model, and an engineering scaling has been applied to the PUMA facility. The counterpart test results indicated that functions of passive safety systems, such as passive containment cooling system (PCCS) start-up, gravity-driven cooling system (GDCS) discharge, PCCS normal operation and overload function were confirmed in both the PANDA and PUMA facilities with qualitative similarities.

Online scenario labeling using a hidden Markov model for assessment of nuclear plant state

By taking into account both aleatory and epistemic uncertainties within the same probabilistic framework, dynamic event trees (DETs) provide more comprehensive and systematic coverage of possible scenarios following an initiating event compared to conventional event trees. When DET generation algorithms are applied to complex realistic systems, extremely large amounts of data can be produced due to both the large number of scenarios generated following a single initiating event and the large number of data channels that represent these scenarios. In addition, the computational time required for the simulation of each scenario can be very large (e.g. about 24h of serial run simulation time for a 4h station blackout scenario). Since scenarios leading to system failure are more of interest, a method is proposed for online labeling of scenarios as failure or non-failure. The algorithm first trains a Hidden Markov Model, which represents the behavior of non-failure scenarios, using a training set from previous simulations. Then, the maximum likelihoods of sample failure and non-failure scenarios fitting this model are computed. These values are used to determine the timestamp at which the labeling of a certain scenario should be performed. Finally, during the succeeding timestamps, the likelihood of each scenario fitting the learned model is computed, and a dynamic thresholding based on the previously calculated likelihood values is applied. The scenarios whose likelihood is higher than the threshold are labeled as non-failure. The proposed algorithm can further delay the non-failure scenarios or discontinue them in order to redirect the computational resources toward the failure scenarios, and reduce computational time and complexity. Experiments using RELAP5/3D model of a fast reactor utilizing an Reactor Vessel Auxiliary Cooling System (RVACS) passive decay heat removal system and dynamic analysis of a station blackout (SBO) event show that the proposed method is capable of correctly labeling 100% of failure scenarios as failure and over 80% of non-failure scenarios as non-failure and provide significant simulation time savings.

Sensitivity analysis of a passive decay heat removal system under a post-loss of coolant accident condition

Passive safety features are now of interest to the design of future generation reactors because of their peculiar characteristics of simplicity, reduction of human interaction, and avoidance of failures of active components. However, the large uncertainty associated with the responses of passive systems might not be ignored. Therefore, it is necessary to identify the uncertain inputs that have the important impact on the uncertainty of the system performance. In this study, two global sensitivity measures, the first-order sensitivity index and the total-order sensitivity index, are applied to a natural circulation decay heat removal system of a gas-cooled fast reactor for identifying the important system inputs. It is found that the uncertainty in the system pressure contributes the most to the uncertainty in the system outputs. In addition, the cooler (the heat exchanger of the emergency cooling system) wall temperature, the Nusselt number in the mixed convection regime, and the friction factor in the mixed convection flow regime also have small impact on the uncertainty of the system outputs.

Decay heat removal in pool type fast reactor using passive systems

Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement of intermediate sodium in SGDHRS with NaK and a decrease in the AHX air inlet temperature do not change the temperatures of the primary circuit significantly. The secondary sodium inventory plays an important role in reducing the temperatures in the primary coolant. The beneficial effect of inter-wrapper flow heat transfer on primary temperatures is limited to about 20K in the fissile zone and 50K in the blanket zone. These results are very important and give direction for future designs of fast breeder reactors.

Reliability estimation of passive systems using fuzzy fault tree approach

In general, a power plant (nuclear, thermal, chemical etc.) consists of operating and emergency safety systems. These systems vary from very complex to simple systems. A system normally consists of active components and passive components. The failure of any operating system will lead to a change in the state of the plant. The availability of the plant depends on the successful operation of the operating systems and the operation of the components in the systems. In order to ensure the availability of the plant reliability of the systems/components should be ensured. In recent years most of the advanced nuclear reactors implement passive systems, aimed at improved safety and availability. In the traditional reliability analysis of passive systems the failure probability is estimated based on the actual components present in the system and their corresponding failure data information. However, the passive system may fail to fulfill its mission not only because of a consequence of classical mechanical failure of component (passive or active) of the passive system, but also due to the deviation from expected behavior due to physical phenomena mainly related to thermal hydraulic (called as virtual component, VC). Hence, one should consider the failure probability of the VC in the analysis. In this paper a methodology for performing passive system reliability, which combines the actual component failures and the failure of VC, has been proposed based on fuzzy fault tree approach. This methodology will eliminate the simulation based approach that is being adopted in the present day passive system reliability analysis. The methodology has been demonstrated with a case study on passive decay heat removal system of a typical nuclear power plant.

Feasibility study on molten gallium with suspended nanoparticles for nuclear coolant applications

Abstract After the Fukushima accident, the concerns on nuclear safety are increasing more than ever before. In particular, promising coolants for fast neutron reactors start to become major targets to the nuclear safety issues. Liquid metals such as sodium, sodium–potassium, lead, and lead–bismuth, as well as gallium have been considered as potential coolants in fast reactors. Among them, gallium is in the liquid state even under room temperature, has the lowest melting point (∼30 °C) and has no explosive reaction with water. However, liquid gallium has a lower thermal conductivity than other liquid metals such as sodium. If nanoparticles were evenly dispersed in liquid gallium, having the higher conductivity the suspended nanoparticles would be expected to produce a highly thermal-conductive coolant for nuclear applications without any safety concerns. At least, it would be promising to adopt the coolant in a safety system like passive decay heat removal system. The present work investigates the feasibility of using liquid gallium with nanoparticles in nuclear applications examining dispersion methods and some of the requirements of nuclear coolants such as absorption cross section, thermal conductivity and viscosity. The key issue of the original idea is whether a stable dispersion can be achieved with various nanoparticles having high thermal conductivities. In the study, the results show that good dispersion stability can be obtained by controlling shear rates based on centrifugal forces.

Multi dimensional analysis of Design Basis Events using MARS-LMR

Abstract KALIMER-600 (Korea Advanced Liquid Metal Reactor), which is a pool type SFR (Sodium-cooled Fast Reactor), was developed by KAERI (Korea Atomic Energy Research Institute). DBE (Design Basis Events) for KALIMER-600 has been analyzed in the one dimension. In this study, the one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions, such as UIS (Upper Internal Structure), IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), and pump. The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. First, the results in normal operation condition show the good agreement between the one and multi-dimensional analysis. However, according to the sodium temperatures of the core inlet, outlet, the fuel central line, cladding and PDRC (Passive Decay heat Removal Circuit), the temperatures of the one dimensional analysis are generally higher than the multi-dimensional analysis in conditions except the normal operation state, and the PDRC operation time in the one dimensional analysis is generally longer than the multi-dimensional analysis. In particular, we can see the reverse and mixing flow phenomena in LOHS better than TOP and LOF, which affect the peak temperature in the one dimensional analysis at 330 s but is not observed in the multi-dimensional analysis. In addition, the different temperatures at the central line of fuel and cladding are shown due to different mass flow in the core. The two PDRCs show the almost same results in one and multi-dimensional analysis, respectively. However, in the comparison between the one and multi-dimension, the heat removal capability of PDRC in the multi-dimensional analysis is better than the one dimensional analysis at the beginning of calculation. The aspects change at the ending of calculation due to the different mass flow in PDRC and over flow sodium temperature. In wrap up, the sodium temperatures of the core in the one dimensional analysis are generally higher than the multidimensional analysis. The temperature peak point in the one dimensional analysis under LOHS does not show in the multi-dimensional analysis, because the reverse and mixing flow pattern between the hot pool and the core outlet region. Even thought there are some different results between the two different dimensional analyses, we can say that the KALIMER-600 system has the safety margins.

Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR) adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS) in design extension conditions (DECs). A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs) and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

PASSIVE SYSTEM OF EMERGENCY REPAIR COOLDOWN

The article examines the system of passive decay heat removal for the emergency repair cooldown of water-moderated water-cooled power reactor (WWRE−1000) in the case of beyond design basis accident with the long-term plant blackout. The description of circuit solutions with the use of heat power equipment on the basis of two-phase thermosiphon is made. The results of calculated stimulation which is done with the help of the thermal hydraulic code MELCOR are also presented.

CFD simulation for steam distribution in header and tube assemblies

The study of the flow characteristics in manifolds (dividing, combining, parallel or Z-manifold and reverse or U-manifold) is a classic subject of engineering fluid dynamics and hydrodynamics. These manifolds are widely used in chemical processes, electronic cooling equipment, solar collectors, spargers, microchannels, fuel cells, heat exchangers and refrigerant distribution in multi-split type of air conditioner, etc. In the literature extensive work has been done for finding out flow distribution in plate-fin heat exchanger, microchannels and spargers. Present work focuses on the flow and pressure distribution in piping networks, which has gained importance in many areas such as air distribution in diffuser system of aerobic biological treatment, steam distribution in passive decay heat removal systems, etc.The uniformity of flow rates among the parallel tubes of piping manifold is governed by the field of fluid pressure in the system under consideration. In present work, the flow and pressure distribution of pure steam in header and tube assemblies has been investigated with the help of CFD simulations. The effects of design parameters has been investigated over a wide range such as the tube pitch (50<Dpt<150mm), header diameter (50<Dh<200mm), tube diameter (15<Dt<32mm), number of tubes (8<Nt<50), inlet or outlet pipe diameter (15<Din<65mm). The sensitivity of inlet mass flow rate (0.033<m˙<0.075 kg/s) has been investigated on the extent of non-uniformity (%ENU) in steam distribution. For validating CFD predictions an experiments have been performed on scaled down geometry of header configuration ‘C1’ using air and water (35,600<Re<68,500) as a working fluid. A good agreement was observed between the predicted and the experimental values of %ENU. It is concluded that the tube diameter, number of tubes and their locational arrangement with respect to inlet and/or outlet pipe are most important design parameters affecting the flow and pressure distribution in the pipeline networks.

Passive system reliability analysis using Response Conditioning Method with an application to failure frequency estimation of Decay Heat Removal of PFBR

Innovative nuclear reactor designs include passive means to achieve high reliability in accomplishing safety functions. Functional reliability analyses of passive systems include Monte Carlo sampling of system uncertainties, followed by propagation through mechanistic system models. For complex passive safety systems of high reliability, Monte Carlo simulations using mechanistic codes are computationally expensive and often become prohibitive. Passive system reliability analysis using recently proposed Response Conditioning Method, which incorporates the insights obtained from approximate solutions like response surfaces in simulations to obtain computationally efficient and consistent probability estimates, is presented in this paper. The method is applied to evaluate the reliability of passive Decay Heat Removal (DHR) system of Indian Prototype Fast Breeder Reactor (PFBR). The accuracy as well as efficiency of the method is compared with direct Monte Carlo simulation. The variability of the reliability values is estimated using bootstrap technique. The system abilities, to prevent critical structural damage as well as to ensure operational safety, are quantitatively ascertained. The system functional failure probabilities are integrated with hardware failure probabilities and the inclusion of passive system unreliability in Probabilistic Safety Assessment is demonstrated.

Validation of CATHARE V2.5 thermal-hydraulic code against full-scale PERSEO tests for decay heat removal in LWRs

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.

PASCAR: Long burning small modular reactor based on natural circulation

PASCAR is a 100 MWt/35 MWe lead–bismuth-cooled small modular reactor which requires no on-site refueling and well suits to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. This paper includes both steady-state and transient performance evaluations for neutronics and thermal–hydraulics. Through design optimization studies for minimizing a burn-up reactivity loss, the metallic fuels-loaded core was designed with less than 1$ reactivity swing over 20-year cycle. A radial peaking power location shows the slow inward migration from outer enrichment zones while maintaining peaking factor within 1.35, reducing radiation damage and corrosion duty of high temperature environments. Equipped with coolant flow path large enough to ensure low pressure drop, this reactor is intended to operate by only natural circulation of chemically inert coolant within relatively low temperature range, 320–420 °C. Peak outlet temperature is nearly 450 °C where an Al-containing duplex cladding has sufficient corrosion resistance. Despite of 50% decrease of fuel thermal conductivity after swelling, inherent negative reactivity feedback and passive decay heat removal capability could secure an ample safety margin of peak fuel centerline temperature in tow safety analyses, unprotected transient overpower and unprotected loss of heat sink. The likelihood of loss of coolant, loss of flow, and local blockage is virtually eliminated by employing respectively a double-walled vessel, pump-less cooling, and cross-flow allowed open square assemblies. Simple fabrication, modular construction, and long burning cycle would compensate for economic disadvantages over smaller power and lower temperature than those of conventional fast reactors.

Failure probability evaluation of passive system using fuzzy Monte Carlo simulation

Passive systems have become an inherent feature of the advanced reactors. The main reason being the passive systems are, theoretically, more reliable than the active ones. Nevertheless the passive system may fail to fulfill its mission not only because of a consequence of classical mechanical failure of component (passive or active) of the passive system, but also due to the deviation from expected behavior due to physical phenomena mainly related to thermal hydraulic or due to different boundary or initial conditions. In this paper the methodology used for performing the passive system reliability analysis has been discussed. A case study on passive decay heat removal system (PDHRS) of large sized pressurized heavy water reactor (PHWR) has been discussed. Thermal hydraulic analysis have been carried out by using RELAP5 code to generate the response surface (from various ranges of identified key parameter values), keeping the criterion as clad surface temperature exceeding certain critical value. Some uncertainties, due to incomplete information, cannot be handled satisfactorily in the probability theory and the fuzzy set theory is more appropriate. In this study the random variables are considered as fuzzy numbers and the fuzzy set theory is employed. In addition, the Monte Carlo simulation technique is utilized to evaluate the probability of failure of system.

Two phase natural convection: CFD simulations and PIV measurement

Buoyancy induced flow and heat transfer are important phenomena in a wide range of engineering systems e.g. electronics and photovoltaics cooling, thermosiphon heat exchangers, solar-thermal heat absorbers, passive decay heat removal systems, etc. Such systems are subject to thermal stratification. The objective of the present work is to study the single phase and two phase (boiling) natural convection accompanied by thermal stratification. We carried out velocity and temperature measurements in a rectangular tank (0.8×0.6×0.6 m 3) fitted with (a) a central tube, and (b) a 10 tube assembly; which form the heat transfer surface. Flows were measured using Particle Image Velocimetry (PIV). Additionally, computational fluid dynamic (CFD) simulations of these systems were performed: first with an assumption of no-boiling (i.e. no phase change) near the heat transfer surfaces; for which we used the open source CFD code OpenFOAM-1.6. For two phase simulations, we used the boiling model of Ganguli et al. (2010) and carried out simulations using the commercial software FLUENT 6.3. The extent of stratification and mixing has been investigated for a range of Rayleigh numbers from 4.34×10 11 to 2.59×10 14. The flow information obtained from PIV was analyzed for insights into the dynamics of turbulent flow structures. We used the signal processing technique of discrete wavelet transform (DWT) for this purpose. From the analysis, we were able to estimate the size, velocity and energy distribution of turbulent structures in our flows. This information was used to estimate wall heat transfer coefficients. A good agreement was observed between the predicted and the experimental values of heat transfer coefficients.

A212 熱流動解析による高温ガス炉圧力容器温度の評価(OS8 高温ガス炉・高速炉・金属燃料)

In the passive decay heat removal system in high-temperature gas-cooled reactor (HTGR), the decay heat is transferred from the reactor pressure vessel(RPV) to cooling panel by radiation and natural convection. The thermal-hydraulic analysis for 1/8 scale mockup experiment simulating passive decay heat removal system of HTGR was performed to demonstrate the numerical calculation accuracy on temperature of RPV. The 3-dimensional 30-deg. sector model which included heater, RPV and cooling panel, was created on analytical code FLUENT v6.3. The temperature on outer face of RPV was numerically calculated within the difference of 10℃ from the experiment data with the condition of helium pressure of 0.47MPa and 0.11MPa in RPV.

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 °C which is substantially lower than ∼627 °C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a crucial factor for consideration in safety design. This study provides future researchers with a guideline on designing safety measures for the fourth generation of the fast reactors with no particular reference to any specific manufacturer.