Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

In this paper, three‐dimensional (3D) power distribution of newly designed small nuclear reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is pressurized water reactor‐based small nuclear reactor in which plate type fuel element has been used and the core of the reactor has hexagonal type geometry. This paper depicts the design of the reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross‐section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the reactor core by NK/TH coupling approach. Copyright © 2012 John Wiley & Sons, Ltd.     

Sub‐channel analysis for Pb‐Bi‐cooled direct contact boiling water fast reactor

Pb‐Bi‐cooled direct contact boiling water fast reactor (PBWFR) featured with a direct‐contact heat exchanger between lead‐bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal‐hydraulic analysis of the PBWFR core in detail. In this paper, a self‐developed SUB‐channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub‐channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub‐channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross‐flow effect would be weakened in the sub‐channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.     

Sub-channel-based multiscale thermal-hydraulic analysis of a medium-power,lead-cooled fast reactor

The lead-cooled fast reactor (LFR) offers enhanced safety and reliability with the fine properties of liquid lead and lead alloy. To study accurately the thermal characteristics of fast reactors, the multiscale thermal-hydraulic coupling simulation is an effective way. Multiscale coupling based on the sub-channel code has evident advantages on the analysis of fuel assemblies. In this study, a multiscale thermal-hydraulic analysis of a forced-circulation, medium-power LFR under steady-state and transient conditions is performed with the system code ATHLET and sub-channel code KMC-SUBtraC which was developed based on the previous version by modifying the pressure drop correlations and adding the assembly-level calculation. The codes are one-way-coupled, with good efficiency and precision. Transient verification of the sub-channel code is conducted with the CFD code. In the steady-state analysis of M²LFR-1000, mass flow and temperature distributions of the assemblies, sub-channels, and fuel rods in the hottest assembly are analyzed and the safety performance is investigated. In the transient analysis, two typical DECs (unprotected overpower transient and ULOF+ULOHS) are simulated and the multiscale thermal-hydraulic characteristics are analyzed. With the negative reactivity feedback, the variations of the temperatures of the coolant and fuel rods are within the safe limits, which shows the inherent safety of the reactor. And the results indicate that the loss of primary flow could increase the risk of cladding corrosion.     

Thermal-hydraulic modeling and transient analysis of pressurized water reactors

This paper examines a new dynamic moving boundary thermal-hydraulic fuel pin model (FUELPIN) for the transient analysis of a pressurized water reactor (PWR). FUELPIN is developed to accommodate the reactor core thermal-hydraulic model of the fuel pin and adjacent coolant flow channel, with detailed thermal conduction in fuel elements. Transient analyses using a known thermal-hydraulic analysis code, COBRA, and FUELPIN linked with a PWR system analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a PWR. The studies of the nuclear reactor system show that moving boundary formulation is highly suitable for the transient thermal-hydraulic analysis of PWRs.     

Thermal‐hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

A 20 MWth, 540 EFPD once through fuel cycle small modular molten salt reactor with solid fuel is proposed by Massachusetts Institute of Technology for off‐grid applications. In this paper, various thermal‐hydraulic analysis methods including computational fluid dynamics, Reactor Excursion Leak Analysis Program (RELAP5), and DAKOTA are adopted step‐by‐step for the reactor design based on the neutronic analysis results. First, 1/12th full core thermal hydraulic analysis is performed by using STAR CCM⁺ with most conservative considerations. Second, the transient safety behaviors of reactor system with risky assumptions are conducted by using REALP5. Finally, due to the unknown factors affecting reactor thermal‐hydraulic characteristics, the uncertainty quantification and sensitivity analysis for the designed reactor is performed with DAKOTA code coupled with RELAP5. Numerical results show that a more uniform temperature distribution with reduced peak temperatures of fuel and coolant across the reactor core has been achieved. Enough safety margin is maintained even under most severe transient accident. The uncertainties in the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. All above results preliminarily indicate the feasibility of the current small modular molten salt reactor design and provide the further optimization direction from reactor thermal‐hydraulic prospective.     

Performance evaluation of Zircaloy reflector for pressurized water reactors

This paper presents detailed analyses of a pressurized water reactor with a new reflector design using zirconium metal. The optimization of the reflector design has been performed using a two‐dimensional fuel assembly reflector model. The three‐dimensional core calculation results with the optimized reflector were compared against those with the existing water reflector and iron reflector. The high scattering cross section of zirconium enhances neutron reflections from the reflector to the core, increasing the peripheral assembly powers. From the analysis based on the equilibrium core, it was noted that the cycle length can be extended, and the pin peaks can be decreased when using zirconium reflector. The analysis has been performed for the optimized power reactor 1000 core with combustion engineering type fuel assemblies using the CASMO‐4E/SIMULATE‐3 (Studsvik Scandpower, Inc., Waltham, MA, USA) code system and SERPENT (VTT Technical Research Centre of Finland, Vuorimiehentie 3, 02150 Espoo, Finland) code, with ENDF/B‐VI data. Copyright © 2015 John Wiley & Sons, Ltd.     

Development of MCBurn and its application in the analysis of SCWR physical characteristics

The MCBurn, a coupled code system linking the Monte Carlo N-particle transport code(MCNP) and Oak Ridge isotope generation and depletion code (ORIGEN), can resolve the basic calculation problems in reactor physical design and analysis, such as criticality, power distribution, nuclide burn up, and neutron fluence. It has been verified in the pressurized water reactor (PWR) pin model, fast reactor (FR) burn up model, and boiling water reactor(BWR) assemble model with benchmarked results. In supercritical water reactor (SCWR) physical calculations, the calculation conditions such as the geometry, the neutron spectrum, and the fuel materials, etc., are more complex than those in traditional reactors, which is a great challenge to reactor physics calculation code. However, the MCBurn code is a possible solution. In this paper, several update functions of the MCBurn in new neutron cross-section lib, code interface, and neutron flux distribution were described. The application of the MCBurn in SCWR were verified on a supercritical water reactor assemble. The calculation results show that the MCBurn is accurate and adaptable in the SCWR calculation.     

Computer simulation of natural convection heat transfer from an assembly of vertical cylinders of PARR-2

In this paper a Computer Code COSINAC (Computer Simulation of Natural Convection from Assembly of vertical Cylinders) has been developed to simulate the natural convection heat transfer from an assembly of vertical cylinders of Pakistan Research Reactor-2 (PARR-2), under the steady state reactor operation. The momentum and energy equations in cylindrical co-ordinates, representing the thermal hydraulic behavior of a typical fuel pin in Pakistan Research Reactor-2, have been solved numerically for a two dimensional axisymmetric domain. The temperature and velocity profiles and Nusselt number variations have been studied and results have been presented. The computer code COSINAC has been validated against experimental results carried out in previous studies at different occasions. Average outlet coolant temperature simulated by computer code, at different wall heat fluxes, has been found in good agreement with experimental results.     

Development of MCBurn and its application in the analysis of SCWR physical characteristics

The MCBurn, a coupled code system linking the Monte Carlo N-particle transport code (MCNP) and Oak Ridge isotope generation and depletion code (ORIGEN), can resolve the basic calculation problems in reactor physical design and analysis, such as criticality, power distribution, nuclide burn up, and neutron fluence. It has been verified in the pressurized water reactor (PWR) pin model, fast reactor (FR) burn up model, and boiling water reactor (BWR) assemble model with benchmarked results. In supercritical water reactor (SCWR) physical calculations, the calculation conditions such as the geometry, the neutron spectrum, and the fuel materials, etc., are more complex than those in traditional reactors, which is a great challenge to reactor physics calculation code. However, the MCBurn code is a possible solution. In this paper, several update functions of the MCBurn in new neutron cross-section lib, code interface, and neutron flux distribution were described. The application of the MCBurn in SCWR were verified on a supercritical water reactor assemble. The calculation results show that the MCBurn is accurate and adaptable in the SCWR calculation.     

Core design of long‐cycle small modular lead‐cooled fast reactor

A core design of small modular liquid‐metal fast reactor (SMLFR) cooled by lead‐bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off‐grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low‐enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S‐CO₂) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the core with low‐enriched uranium (LEU) performs the function of core ignition. The center region plays the role of a breeding blanket to increase the core lifetime for long cycle operation. The core working fluid inlet and outlet temperatures are 300°C and 422°C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi‐static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal‐hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the conceptual core is favorable in terms of self‐controllability, which is the first step towards inherent safety.     

Preliminary design of a medium‐power modular lead‐cooled fast reactor with the application of optimization methods

Based on research and development experience from Gen III, Gen III+, and Gen IV reactor concepts, a 1000‐MWt medium‐power modular lead‐cooled fast reactor M²LFR‐1000 was developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing application of optimization methods in preliminary design phase. By using the optimization methods presented, primarily considering the safety design limits (the maximum coolant velocity, the maximum cladding temperature, and the maximum burn‐up limited by the cladding radiation damage permitted), the preliminary design of 1000‐MWth medium‐power modular lead‐cooled fast reactor M²LFR‐1000 was carried out, including the design of fuel rods, fuel assemblies, reactivity control system, primary system, secondary system, decay heat removal system, and so on. The analysis of neutron characteristics (including reactivity feedback coefficients) and thermal hydraulics characteristics (the maximum fuel temperature and the maximum cladding outer surface temperature) of the core under normal steady‐state condition was carried out to evaluate the core design. Also, the analysis of 2 typical protected transients (protected transient over power accident and protected loss of flow accident) was conducted. Other analysis work of the reactor is to be done, such as the transient analysis via computational fluid dynamic codes and the seismic response analysis of the reactor. But the preliminary analysis results obtained so far under normal steady state and transient conditions confirm the inherent safety characteristics of the reactor design.     

CFD analysis of heat transfer in hexagonal subchannels of super fast reactor in upward flow

A super fast reactor is a fast spectrum, supercritical, water‐cooled reactor. This paper represents CFD analysis of heat transfer in hexagonal subchannels of super fast reactor using FLUENT in ANSYS. The numerical simulation of grid stability was done by considering different mesh sizes and the turbulence model for heat transfer of supercritical water was also carried out and compared with the experimental data. RNG k‐? turbulence model with enhanced wall treatment was considered for simulations. Heat transfer and heat generation rate analysis of the outer surface rod wall is carried out with different subchannels by changing various parameters like boundary conditions and pitch‐to‐diameter ratio. The analyses reveal that the outer surface of the rod wall temperature decreases with increase in pitch‐to‐diameter ratio. Maximum coolant temperature rises in edge subchannels more than corner subchannels. Further analysis is carried out with different mass fluxes. Increases in mass flux has minimal effect on the maximum rod wall surface temperature. Maximum cladding surface temperature for the corner subchannel is less compared to the edge subchannel. Heat generation rate also decreases with increase in pitch‐to‐diameter ratio. This paper also investigates the buoyancy effect on subchannels with varying heat flux as boundary conditions considering constant mass flux.     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR ZPR ‐ LFR analysis

Studies related to severe core accidents constitute a crucial element in the safety design of Gen‐IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero‐power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen‐IV cores by using critical facilities and surrogate models. The current study introduces the European lead‐cooled system (ELSY) as an additional Gen‐IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium‐cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR fast/thermal coupled ZPRs. Part I—Assembly level

The comprehension of severe criticality accident is a key issue in Gen‐IV neutronics and safety. Within the future zero‐power experimental physics reactor (ZEPHYR), to be built in Cadarache in the next decade, innovative approaches to reproduce high temperature partially degraded Gen‐IV cores into a critical facility is being investigated. This work presents the first attempt to represent a fuel assembly of sodium‐cooled fast reactor severe criticality accident based on surrogate models. One identified way to construct such representative configuration is to use MASURCA plates stockpile (MOX, UOx, Na, U, and Pu metal) in a fast/thermal coupled core to model a stratified molten assembly. The present study is the first step in a more global approach to full core analysis. The approach is based on a nature‐inspired metaheuristic algorithm, the particle swarm optimization algorithm, to find relevant ZEPHYR configuration at 20°C that exhibits characteristics of (2000‐3000°C) molten MOX assembly in a stratified metal arrangement in a reference sodium‐cooled fast reactor core. Thus, the underlying research question of this study is whether it is possible to represent temperature‐related reactivity effects occurring at fuel meltdown temperatures in a power reactor as density‐related reactivity effects at the operation temperature of a zero‐power reactor, and if so, how should it be done? The calculations are based on a Serpent‐2 Monte Carlo sensitivity and representativity analysis using the Cadarache's cross sections covariance data (COMAC). The single fuel assembly studies show that it is possible to represent the multiplication factor with a representativity factor greater than 0.98. As for reactivity variations, it is possible to achieve a satisfactory representativity factor of above 0.85 in all the presented cases. The representativity process demonstrates that temperature effects could be translated into density effects with good confidence. A complementary analysis on modified nuclear data covariance matrix demonstrates the importance of selecting consistent and robust uncertainties in the particle swarm optimization algorithm. This work provides insights on the behavior of the representativity scheme in different core states and shades some light on the problem in hand.     

Performance Evaluation of Two Heat Transfer Models of a Walking Beam Type Reheat Furnace

Two different heat transfer models for predicting the transient heat transfer characteristics of the slabs in a walking beam type reheat furnace are compared in this work. The prediction of heat flux on the slab surface and the temperature distribution inside the slab have been determined by considering thermal radiation in the furnace chamber and transient heat conduction in the slab. Both models have been compared for their accuracy and computational time. The furnace is modeled as an enclosure with a radiatively participating medium. In the first model, the three-dimensional (3D) transient heat conduction equation with a radiative heat flux boundary condition is solved using an in-house code. The radiative heat flux incident on the slab surface required in the boundary condition of the conduction code is calculated using the commercial software FLUENT. The second model uses entirely FLUENT along with a user-defined function, which has been developed to account for the movement of slabs. The results obtained from both models have a maximum temperature difference of 2.25%, whereas the computational time for the first model is 3 h and that for the second model is approximately 100 h.     

Analysis of load‐following characteristics for an integrated pressurized water reactor

Integrated pressurized water reactor (IPWR) usually be equipped with once‐through steam generators (OTSGs). The OTSG has many advantages such as simple mechanical structure, smaller size, and higher heat transfer efficiency. It produces superheated steam but with less inventory in its secondary side. The steam pressure is easily affected by steam flow rate or feed water flow rate. This draws more attention to design advanced reactor control system. In this paper, a study has been carried out to analyze the thermal hydraulic performance of an advanced IPWR under steady‐state and transient conditions by using a thermal hydraulic safety analysis code Relap5. An effective load‐following control system is proposed. The steady‐state operating characteristics of IPWR at different load conditions show that the average primary coolant temperature, steam pressure, and coolant mass flow rate are the most important control parameters. Pump frequency conversion strategy and OTSG grouping run strategy are used to study the transient operating characteristics of IPWR. Simulation results of the control system demonstrate its capability in regulating feedwater flow rate and reactor power to follow the change of steam flow rate. According to the results, the OTSG grouping run strategy is optimized to ensure the OTSG operates safely under low load conditions. Copyright © 2013 John Wiley & Sons, Ltd.     

Conceptual design of an innovative reduced moderation thorium‐fueled small modular reactor with heavy‐water coolant

This paper presents an innovative conceptual design for small modular reactors, the reduced‐moderation small modular reactor (RMSMR), for the sustainable use of nuclear resources. The concept is established by a modification of the well‐understood pressurized water reactor technology. A reduced‐moderation lattice and heavy‐water coolant are used to yield an epithermal‐to‐fast neutron spectrum, which is beneficial for attaining a large conversion ratio and reducing the burnup reactivity swing throughout the core lifetime. Two‐dimensional pin cell and three‐dimensional core burnup calculations are performed to systematically analyze the neutronics influences of important parameters, such as the coolant type, moderator‐to‐fuel ratio, and fuel type. The RMSMR adopts a three‐zone uranium‐thorium dioxide fuel configuration to flatten the power distribution and ensure a negative void coefficient. The radial and axial blanket regions are found to enhance the breeding effect. The proposed RMSMR can sustain power generation of 100 MWe for 7 years without refueling and achieve a conversion ratio of 0.85 at the end of the cycle. Numerical simulations indicate that the proposed concept has satisfactory shutdown margins and reactivity coefficients and conservative thermal‐hydraulic safety. The RMSMR may be a promising candidate to fill the gap between light‐water reactors and fast breeder reactors.     

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

Lead‐based fast reactors (LFRs) have unique advantages in the development of a SMR, which has attracted a lot of attention in recent years. In this paper, an optimized design for a lead‐bismuth small modular reactor was studied on the basis of the design of SUPERSTAR. This paper aims to propose an improved LFR core scheme to enhance the neutronic performance as well as the thermal‐hydraulic safety of the reference reactor. Advanced nitride fuel is adopted in which the plutonium is used as the driven fuel, while thorium is used as the fertile fuel. Subchannel analysis was performed in the assembly design using an in‐house subchannel code, SUBAS, and an 11 × 11 scheme with a pitch‐to‐diameter (P/D) ratio of 1.4 was chosen. Using the modified assembly, the core was redesigned using the coupled code MCORE. The active core was divided into four zones with different enrichment of ²³⁹Pu to extend the core lifetime and flatten the power distribution. The main kinetic parameters and reactivity coefficients were obtained. Neutronic performance at different operation times was also studied. The maximum radial power peak factor was 1.28, while the maximum total power peak factor was 1.737. During the whole lifetime, the reactivity swing was 0.926$, which was below the limit of 1$. The subchannel study of the core flow distribution showed that a flow distributor is needed to further improve the flow distribution capability. The peaking cladding temperature was 508.7°C, and the maximum fuel center temperature was 723.4°C, both of which do not exceed the limit temperature. Compared with features of SUPERSTAR, the peaking cladding temperature was well improved and the lifetime extended.     

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Partitioning and transmutation of the minor actinides (MAs) from nuclear power plants are of importance for the nuclear‐energy sustainable development. Fast reactors are applied to transmutation for the hard neutron spectrum and high neutron flux. However, the safety‐related neutronic parameters will become worse when large amounts of MAs are loaded. In this paper, transients of a 600 MWe sodium‐cooled fast reactor for MA transmutation are analyzed by using neutron transport simulation. The control rod withdrawal transients are calculated. Two cases are compared to investigate the influence of loading MAs into the fast reactor core. One is the common core loaded with mixed oxide (MOX) fuel, and the other one is the transmutation core loaded with MOX fuel and MAs. The results indicate that in order to apply similar operation criterion with the common core, the transmutation core with 6% weight fraction of MAs should be operated with more than 30% power reduction. In addition, the results of the transport‐based transient analysis and the point kinetics transient analysis are compared. There are noticeable differences between them, which indicate that the usual way based on the point kinetics calculation is not suitable well for simulating the control rod introduced transients. Copyright © 2017 John Wiley & Sons, Ltd.     

Conceptual study of an accelerator‐driven ceramic fast reactor with long‐term operation

To improve both safe operation and high resource utilization in nuclear power, we propose and investigate the concept of an accelerator‐driven ceramic fast reactor (ADCFR). This reactor type has the potential to operate continuously throughout a 40‐year core life, without fuel shuffling or supplementation. The ADCFR consists of a high‐power superconducting linear accelerator, a gravity‐driven dense granular spallation‐target, and a ceramic fast reactor. The performance of the ADCFR was assessed by using a neutron‐physics simulation, thermal calculations, and a characteristic analysis. The results show that the peak position for the neutron spectrum in the ADCFR is at about 0.1 MeV. This means that it falls with the fast neutron spectrum, and it can convert loaded nuclear fertile material into fissile fuel. Using a burnup simulation, the ideal effective multiplication‐factor (K_eff) was calculated by using a combination of subcritical (accelerator‐driven) and critical modes. In 40 year of operation, K_eff is obtained from the initial 0.98 to the peak ~1.02 and then to ~0.99. Different granular coolant materials were selected to compare neutron performance. In breeding, the differences are relatively small. The thermal calculation indicates that heat transfer performance of granular makes it possible to meet the required specifications in theory. Finally, the corresponding characteristics, with regard to the 2‐phase coolant, ceramic materials, nuclear safety performance, operation modes, economics, and range of applications were analyzed. Accelerator‐driven ceramic fast reactors can achieve very high levels of inherent safety, good breeding performance, high power‐generation efficiency, and high flexibility in wide range of applications.