Adjoint-based sensitivity analysis of circulating liquid fuel system for the multiphysics model of molten salt reactor

The strongly coupled behaviors between neutronics and thermal-hydraulics of liquid-fueled molten salt reactors make it difficult to evaluate system behaviors, due to the transport of precursors along moving fuel. Extending an adjoint-based method on the multiphysics approach, different assumptions on temperature dependencies of nuclear and thermophysical properties of salt are included in the local sensitivity analysis of a circulating liquid fuel system. Local sensitivity of various types of system response in steady-state is analyzed for 39 parameters including coupling models, reactor design values, and kinetic constants of delayed neutron and decay heat precursors for a simplified 1D model of molten salt fast reactor. Extended adjoint-based sensitivity analysis method for MSR is successfully validated achieving 1.38% deviation on average between a recalculation and adjoint method, comparing local sensitivities to all parameters. Also, it takes 66.3 times less in computational time compared with the recalculation method for evaluating the sensitivity of the same type of system response. The importance of all the parameters to the system response is analyzed according to the assumptions on temperature dependencies to nuclear data and salt properties. The most influencing ones are fission energy-related terms, and their importance increases when temperature dependencies are taken into account, compared with constant properties. Changes of influences on the sensitivity are investigated from the relative changes of the parameter values in various system response types, and it implies the importance to consider the multiphysics modeling on the local sensitivity analysis.     

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.     

Thorium utilization in a small modular molten salt reactor with progressive fuel cycle modes

The thorium‐uranium (Th‐U) fuel cycle is considered as a potential approach to ensure a long‐term supply of nuclear fuel. Small modular molten salt reactor (SMMSR) is regarded as one of the candidate reactors for Th utilization, since it inherits the merits of both MSR and small modular reactor. The Th utilization in a 220‐MWe SMMSR with the once‐through fuel cycle mode is investigated first. Then, the SMMSR with batch and online fuel processing modes is investigated second for comparison, considering the progressive development of fuel reprocessing technology. To keep a negative temperature reactivity feedback coefficient (TRC), a configuration for fuel salt volume fraction (SVF) equal to 15%, with a mixed fuel of low enriched uranium (LEU) and thorium at an operation time of 5 years is recommended for the once‐through mode, corresponding to the Th energy contribution (ThEC) of 37.6% and natural U and Th utilization efficiency (UE) of 0.51%. Considering the solubility limit of heavy nuclide (HN) proportion (below 18.0 mol%) in the fuel salt, the total operation time of the SMMSR shall be less than 50 years for the batch reprocessing mode with a 5‐year reprocessing interval time. In this case, the ThEC and UE can be improved to about 47.4% and 0.99%, respectively. Finally, the Th utilization and fuel sustainability are analyzed at a lifetime of 50 years for the online reprocessing fuel cycle mode, including both the only online fission products (FPs) removing scheme and the fuel transition scheme from LEU to ²³³U. For the former scheme, the ThEC and UE can be further improved to 58.6% and 1.52%, respectively. For the latter scheme, ²³³Pa is extracted continuously from the core to breed and store ²³³U. If a total reactor lifetime of 50 years is assumed, the operation time using LEU as starting and feeding fuel for 6 years is required, and the bred ²³³U during this 6‐year operation can start and maintain the reactor criticality for the remaining 44 years. In this case, the ThEC is improved significantly to 89.1% corresponding to a UE of 2.74%.     

Experimental study on the heat transfer characteristics of fluoride salt in the new conceptual passive heat removal system of molten salt reactor

A new conceptual design of a passive residual heat removal system (PRHRS) has been proposed for molten salt reactor. High‐temperature heat pipes are used in this new design to improve the system inherent safety and make the PRHRS more compact. An experimental system using fluoride salt FLiNaK has been constructed to validate and support the future design of PRHRS of molten salt reactors. In this research, tests on the natural convection heat transfer of FLiNaK in the drain tank with an inclined heat pipe inserted at different heights were performed. The temperature distribution of fluoride salt in the tank was analyzed. The height of heat pipe and the bulk temperature of FLiNaK have little influence on the normalized salt temperature distribution. However, with the height of heat pipe increasing, the temperature difference of molten salt decreases and heat transfer coefficient of natural convection increases. In addition, the empirical correlations of natural convection heat transfer between liquid FLiNaK and inclined heat pipe are obtained within the range of Rayleigh numbers from 3.97 × 10⁶ to 1.16 × 10⁷. The comparisons show that a good agreement with less than 5% deviation is obtained between the proposed correlations and the test data.     

Transient characteristics of small molten salt reactor during blockage accident

This paper reports the results from a transient core analysis of a small molten salt reactor (MSR) when a duct blockage accident occurred. The focus of this study is a numerical model employed in order to consider the interaction among fuel salt flow, heat transfer, and nuclear reactions. The numerical model comprises continuity and momentum conservation equations for fuel salt flow, two‐group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six‐group delayed neutron precursors, and energy conservation equations for fuel salt and graphite moderators. The analysis results show the following: (1) the effect of the self‐control performance of the MSR on the effective multiplication factor and thermal power output of the reactor after the blockage accident is insignificant, (2) fuel salt and graphite moderator temperatures increase drastically but locally at the blockage area and its surroundings, (3) the highest fuel salt temperature after the blockage accident is 1,363 K; this value is lower than the boiling point of fuel salt and the melting temperature of the reactor vessel, (4) the change in the distributions of fast and thermal neutron fluxes after the blockage accident when compared with the distributions at the rated condition is very slight, and (5) delayed neutron precursors, especially the first delayed neutron precursor, accumulate at the blockage area due to its large decay constant. These results imply that the safety of the MSR is assured in the case of a blockage accident. © 2006 Wiley Periodicals, Inc. Heat Trans Asian Res, 35(6): 434–450, 2006; Published online in Wiley InterScience ( www.interscience.wiley.com ). DOI 10.1002/htj.20123     

Review of conceptual design and fundamental research of molten salt reactors in China

Molten salt reactor (MSR) as 1 candidate of the generation IV advanced nuclear power systems attracted more attention in China due to its top ranked in fuel cycle and thorium utilization. Two types of MSR concepts were studied and developed in parallel, namely the MSR with liquid fuel and that with solid fuel. Abundant fundamental research including the neutronics modeling, thermal‐hydraulics modeling, safety analysis, material investigation, molten salts technologies etc. were carried out. Some analysis software such as COUPLE and FANCY were developed. Several experimental facilities like high‐temperature fluoride salt experiment loop have been constructed. Some passive residual heat removal systems were designed, and 1 test facility is under construction. The key MSR techniques including the extraction and separation of molten salt and construction of N‐base alloy have been mastered. Based on these fundamental research, Chinese Academy of Sciences has completed the design of thorium‐based MSRs with solid fuel and liquid fuel and is promoting their construction in the near future. In China, future efforts should be paid to the material, online fuel processing, Th‐U fuel cycle, component design, and construction and thermal‐hydraulic experiments for MSR, which are rather challenging nowadays.     

Thermal‐hydraulic design features of a micronuclear reactor power source applied for multipurpose

Microheat pipe cooled reactor power source (HRP) designed for space or underwater vehicles meets the future demands, such as safer structure, longer operating time, and fewer mechanical moving parts. In this paper, potassium heat pipe cooled reactor power source system which generates 50 kWe electricity is proposed. The reactor core using uranium nitride fuel is cooled by 37 potassium high‐temperature heat pipes. The shields are designed as tungsten and water, and reactor reactivity is controlled by control drums. The thermoelectric generator (TEG) consists of thermoelectric conversion units and seawater cooler. The thermoelectric conversion units convert thermal energy to electric energy through the high‐performance thermoelectric material. A code applied for designing and analyzing the reactor power system is developed. It consists of multichannel reactor core model, heat pipe model using thermal resistance network, thermoelectric conversion, and thermal conductivity model. Then, the sensitivity analysis is performed on two key parameters including the length of the heat pipe condensation section and the cold junction temperature of the TE cell. Meanwhile, the steady‐state calculations are conducted. Results show that the maximum fuel temperature is 938 K located in the center of reactor core and the outlet temperature of coolant reaches 316 K. Both of them are within the limitation. It is concluded that the preliminary design of HPR design is reasonable and reliable. The designed residual heat removal system has sufficient safety margin to release the decay heat of the reactor. This research provides valuable analysis for the application of micronuclear power source.     

Control‐oriented nonlinear modelling of molten carbonate fuel cells

Performance and availability of molten carbonate fuel cells (MCFC) stack are greatly dependent on its operating temperature. Control of the operating temperature within a specified range and reduction of its temperature fluctuation are highly desirable. The models of MCFC stack existing are too complicated to be suitable for design of a controller because of its lack of clear input–output relations. In this paper, according to the demands of control design, a quantitative relations model of control‐oriented MCFC between the temperatures of the stack and flowrates of the input gases is developed, based on conservation laws. It is an affine nonlinear model with multi‐input and multi‐output, the flowrates of fuel and oxidant gases as the manipulated vector and the temperatures of MCFC electrode–electrolyte plates, separator plates as the controlled vector. The modelling and simulation procedures are given in detail. The simulation tests reveal that the model developed is accurate and it is suitable to be used as a model in designing a controller of MCFC stack. Copyright © 2004 John Wiley & Sons, Ltd.     

Pretest in forced circulation molten salt heat transfer loop: Studies with thermia‐B

Molten salts have potential application as an efficient heat transfer medium in a primary and secondary heat exchanger in high temperature next‐generation nuclear power plant. Thermal hydraulic studies are vital for reliable and cost‐effective design of the nuclear power plant. Therefore heat transfer study of molten salts will play a vital role in this area. In this work, an experimental system was designed to study thermal hydraulics of the molten salt system up to 700°C. This work describes the pretest results of the experimental facility for extremely corrosive molten fluoride salts with a simulant thermia‐B as the working fluid. In the present work, the details of the system are discussed and thermal‐hydraulic data for heat transfer fluid thermia‐B has been presented. Experiments were carried out at Reynolds number in the range of 4500 to 40 500 and Prandtl number in the range of 34 to 144. Effect of Reynolds number, melting tank temperature, and heat input to test section on forced convective heat transfer was studied under turbulent conditions. Comparison of the experimental data with different empirical correlations has been presented.     

Plate flat heat pipe and liquid‐cooled coupled multistage heat dissipation system of Li‐ion battery

A thermal management system with the capability of achieving excellent heat dissipation is essential to the development of battery pack for transportation devices. To meet the temperature uniformity requirements of the battery pack, the plate flat heat pipe and liquid‐cooled coupled multistage heat dissipation system had been introduced. In this article, the research status of thermal management systems in battery pack was introduced. And the heat generation and heating power of the Li‐ion cell were studied. Then, the structure model of plate flat heat pipe system was proposed. Finally, the enhanced heat conduction effect of the thermal management system proposed in this article was comprehensively analyzed. Through the analysis of the results, in high discharge rates, the thermal management system proposed in this article could meet the temperature uniformity requirements of battery pack; also, the internal difference would reduce by 30.20%.     

Determination of the liquid eutectic metal fuel dual fluid reactor (DFRm) design – steady state calculations

The dual fluid reactor (DFR) is a novel concept of a very high‐temperature (fast) reactor that falls off the classification of generation 4 international forum (GIF). DFR makes best of the two previous designs: molten salt reactor (MSR) and lead‐cooled fast reactor (LFR). In this paper, we present a new reactor design Dual Fluid Reactor metallic (DFRm) with the liquid eutectic U‐Cr metal fuel composition and the lead coolant of which general idea was patented recently. By performing the first steady state neutronic calculations for such a reactor (the neutron flux density as a function of energy, the burnup, the effective multiplication factor/reactivity), we show that this 250‐MWth reactor is critical, and that it can operate almost 20 years without refuelling. We also optimise the geometry (reflector thickness, fuel tubes pin pitch) with respect to the multiplication factor. The optimisation together with some other opportunities for the liquid metal fuel design (eg, the use of electromagnetic pumps to circulate the medium) allows DFRm to be of a small size. This rises economy of the construction as expressed nicely in terms of the energy return on invested (EROI) factor, which is even higher than for the molten salt fuel design (DFRs). Last but not least, we show that DFRm has all the (fuel, coolant, reflector) temperature coefficients negative, which is an important factor of the passive safety.     

Oil fouling in double‐pipe heat exchanger under liquid‐liquid dispersion and the influence of copper oxide nanofluid

Dispersions of oil in water are encountered in a variety of industrial processes leading to a reduction in the performance of the heat exchangers when thermally treating such two phase fluids. This reduction is mainly due to changes in the thermal and hydrodynamical behavior of the two phase fluid. In the present work, an experimental investigation was performed to study the effects of light oil fouling on the heat transfer coefficient in a double‐pipe heat exchanger under turbulent flow conditions. The effects of different operating conditions on the fouling rate were investigated including: hot fluid Reynolds number (the dispersion), cold fluid Reynolds number, and time. The oil fouling rate was analyzed by determining the growth of fouling resistance with time and through pressure drop measurements. The influence of copper oxide (CuO) nanofluid on the fouling rate in the dispersion was also determined. It was found that the presence of dispersed oil causes a reduction in heat transfer coefficient by percentages depending on the Reynolds number of both cold and hot fluids and the concentration of oil. In addition, the time history of fouling resistance exhibited different trends with the flow rates of both fluids and its trend was influenced appreciably by the presence of CuO nanofluid.     

Power flattening study of ultra‐long cycle fast reactor using thorium fuel

This paper presents a design study of power shape flattening for an optimized ultra‐long cycle fast reactor with a power rate of 1000 MWe in order to mitigate the power peaking issue and improve the safety with a lower maximum neutron flux and reactivity swing. There are variations in the core designs by loading thorium fuel or zoning fuels in the blanket region and the bottom driver region of ultra‐long cycle fast reactor with a power rate of 1000 MWe. While it has lower breeding performance in a fast breeder reactor, thorium fuel is one of the promising fuel options for future reactors because of its abundance and its safety characteristics. It has been confirmed that the thorium fuels, when loaded into the center region of a reactor core, lower the power peaking factor from 1.64 to 1.25 after 20 years and achieves a more flattened radial power distribution. This consequently reduces the maximum neutron flux and the speed of the active core moving from 3.0 cm/year to 2.5 cm/year on the average over the 60‐year reactor operation. It has been successfully demonstrated that the three‐zone core is the most optimized core, has the most flattened radial power shape, and is without any compromise in the nature of long cycle core, from the neutronics point of view, in terms of average discharge burnup and breeding ratio. Copyright © 2016 John Wiley & Sons, Ltd.     

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.     

Core design options of an ultra‐long‐cycle sodium cooled reactor with effective use of PWR spent fuel for sustainable energy supply

In this work, 350MWe ultra‐long‐cycle sodium‐cooled reactor cores are designed to supply electric energy over ~60 Effective Full Power Years (EFPYs) without refueling and with an effective use of Transuranics (TRU) and uranium from large pressurized water reactor (PWR) spent fuel stocks. The core employs the axial blanket‐driver‐blanket (ABDB) burning strategy, which was recently proposed by the authors to achieve an ultra‐long‐cycle length with self‐controllability under unprotected accidents. In particular, a thorium–uranium fuel cycle is considered to remove the heterogeneity of the fuel assemblies for design simplification and to improve the core performance parameters by selectively adding thorium into both blanket and driver fuels. The results show that the use of TRU nuclides from PWR spent fuel leads to significant extension of the fuel cycle length, but considerable increase of burnup reactivity swing. In addition, these results also indicate that the uranium–thorium mixed fuels both in the lower blanket and driver considerably improve the inherent safety of the ultra‐long‐cycle core by reducing burnup reactivity and sodium void worth; this makes it possible to simplify the previous heterogeneous fuel assembly design with improved core performances. Copyright © 2016 John Wiley & Sons, Ltd.     

Experimental and numerical analysis of oxy‐fuel combustion in a porous plate reactor

The present work focuses on studying experimentally and numerically the oxy‐fuel combustion characteristics inside a porous plate reactor towards the application of oxy‐combustion carbon capture technology. Initially, non‐reactive flow experiments are performed to analyze the permeation rate of oxygen in order to obtain the desired stoichiometric ratios. A numerical model is developed for non‐reactive and reactive flow cases. The model is validated against the presently recorded experimental data for the non‐reacting flow cases, and it is validated against the available literature data for oxy‐fuel combustion for the reacting flow cases. A modified two‐step oxy‐combustion reaction kinetics model for methane is implemented in the present model. Simulations are performed over wide range of operating oxidizer ratios (O₂/CO₂ ratio), from OR = 0.2 to OR = 0.4, and over wide range of equivalence ratios, from φ = 0.7 to φ = 1.0. The flame length was decreased as a result of the increase of the oxidizer ratio. Effects of CO₂ recirculation amount on the oxy‐combustion flame stability are examined. A reduction in combustion temperature and increase in flame fluctuations are encountered while increasing CO₂ concentration inside the reactor. At high equivalence ratio, the combustion temperature and flame stability are improved. At low equivalence ratio, the flame length is increased, and the flame was moved towards the reactor center line. Copyright © 2015 John Wiley & Sons, Ltd.     

Non‐isothermal multi‐phase modeling of PEM fuel cell cathode

In this study, numerical simulation has been carried out for the heat transfer and temperature distribution in the cathode of polymer electrolyte membrane fuel cells along with the multi‐phase and multi‐species transport under the steady‐state condition. The commercial software, COMSOL Multiphysics, is used to solve the conservation equations for momentum, mass, species, charge and energy numerically. The conservation equations are applied to the solid, liquid and vapor phases in the bipolar plate and gas diffusion (GDL) and catalyst layers of a two‐dimensional cross section of the cathode. The catalyst layer is assumed to be a finite domain and the water production in the catalyst layer is considered to be in the liquid form. The temperature distribution in the cathode is simulated and then the effects of the relative humidity of the air stream, the permeability of the cathode and the flow channel shoulder to channel width ratio are investigated. It is shown that the highest temperature change, both in the in‐plane and across‐the‐plane directions, occurs in the GDL, while the highest temperature is reached in the catalyst layer. The distribution of temperature in the bipolar plate is shown to be relatively uniform due to the high thermal conductivity of the plate. A decrease in the inlet relative humidity of the air stream results in the decrease of the maximum temperature due to the absorption of heat during the evaporation of liquid water in the GDL and catalyst layer. The non‐uniformity of the temperature distribution, especially in the catalyst layer, is observed with the increase of the permeability of the cathode. Similarly, the decrease of the channel shoulder to channel width ratio leads to a non‐uniform distribution of temperature especially under the channel areas. Copyright © 2009 John Wiley & Sons, Ltd.     

Electrical and electrical–thermal power plants with molten carbonate fuel cell/gas turbine‐integrated systems

In this article the calculation tool further developed and implemented in Matlab language by the authors was used to determine some optimal operating conditions in electrical and thermal or electrical terms for two different types of hybrid systems: molten carbonate fuel cells (MCFC)/gas turbine with their heat recovery system and the hybrid systems operating in these optimal conditions were analyzed. In the heat recovery system, in both cases, a part of the thermal energy of these gases is used to produce the steam necessary for the MCFC system. The remaining thermal energy is used, in one case, for the production of steam at various levels of pressure and temperature, which feeds a steam bottom plant to produce additional electric energy; in the other case, the same thermal energy is used to produce steam for cogenerative use. The heat recovery system was suitably designed according to the circumstances and the performances and the specific CO₂ emissions of the hybrid systems were evaluated. Copyright © 2010 John Wiley & Sons, Ltd.     

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.