Fast Neutron Reactors Research Articles

Measurement of γ-Rays Generated by Neutron Interaction with 16O at 30 MeV and 250 MeV

Abstract Deep understanding of $\gamma$-ray production from the fast neutron reaction in water is crucial for various physics studies at large-scale water Cherenkov detectors. We performed test experiments using quasi-mono energetic neutron beams ($E_n = 30$ and 250 MeV) at Osaka University’s Research Center for Nuclear Physics to measure $\gamma$-rays originating from the neutron–oxygen reaction with a high-purity germanium detector. Multiple $\gamma$-ray peaks which are expected to be from excited nuclei after the neutron–oxygen reaction were successfully observed. We measured the neutron beam flux using an organic liquid scintillator for the cross section measurement. With a spectral fitting analysis based on the tailored $\gamma$-ray signal and background templates, we measured cross sections for each observed $\gamma$-ray component. The results will be useful to validate neutron models employed in ongoing and future water Cherenkov experiments.

Determination of the silicon content of geological materials by reactor fast neutron activation analysis

Determination of the silicon content of geological materials by reactor fast neutron activation analysis

A self-organized sandwich structure of chromium nitride for ultra-long lifetime in liquid sodium.

The development of fast neutron reactors with improved efficiency and sustainability, being a tangible solution to the large-scale utilization of nuclear energy, serves as a critical step prior to the commercialization of fusion energy. These reactors use liquid metal coolants, which can weaken the durability of metallic components. Conventional design of protective coatings counts upon thermodynamics, which often overlooks the kinetic factors such as structural evolutions, resulting in deteriorated coating properties. Herein, we present a novel interface-engineering strategy involving the control of the phase transformation direction and interface diffusion reaction. Through iterations of self-organization, desired surfaces and interfaces can be achieved for materials used in harsh environments. Specifically, a CrN-coated steel sample with an interfacial Cr layer was designed and fabricated. After ultra-long (up to 6000 h) immersion in liquid sodium, the CrN/Cr coating structure was converted into a sandwich Cr2N/CrN/Cr2N structure dynamically. As a consequence, the coating system exhibited enhanced properties, namely increased surface hardness (by ∼36%), reduced coefficient of friction (by ∼13%), and enhanced interfacial adhesion (by ∼37%). Thus, the proposed strategy can guide the future design of robust coatings with ultra-long service life in harsh environments.

Simulation Analysis Based on Sodium-Cooled Fast Reactor Fuel Assembly Under Blockage Accident

This paper employs computational fluid dynamics to investigate the fuel assemblies within a sodium-cooled fast neutron reactor. To begin with, we developed computational models for seven wire spacer fuel rods under both normal operating conditions and transient blockage conditions. We conducted separate analyses to assess the impacts of normal operation, blockage thickness, and blockage area. This work allowed us to acquire data on the heat transfer properties and the flow field variations for the coolant, blockages, and fuel rods across different conditions. Subsequently, we leveraged these flow field alterations to examine the resulting temperature distributions. Analyses were conducted to evaluate the effects of normal operation, blockage thickness, and blockage area. The research acquired the heat transfer characteristics and flow field distribution variations of the coolant, blockages, and fuel rods under different operational conditions and utilized these variations to analyze the temperature distribution. Through research analysis, the following conclusions have been reached. Under normal operating conditions, the temperature and flow fields of the fuel components exhibit cyclic variations along the axial length, corresponding to the pitch of the wire spacers. Heat exchange between the internal and external subchannels occurs independently, which further substantiates that the incorporation of wire spacers strengthens lateral flow disturbances. This effect is more pronounced within the internal subchannels, thereby leading to a marked difference in the flow fields on either side of the wire spacers. In the case of blocked conditions, an increase in blockage thickness and area both lead to higher temperatures for the coolant, the blockages themselves, and the fuel rods. The temperature in the recirculation zone behind the blockages also rises with increasing blockage thickness and area, although the magnitude of this increase is not significant. After the onset of blockage conditions, the instantaneous temperature tends to increase. As time progresses, the instantaneous temperature fields following the blockage do not display greater fluctuations in temperature change than those observed after the system has stabilized. For the temperature parameters of the convective field, differences arising from an increase in blockage area are more significant than those caused by an increase in blockage thickness. When blockage area and blockage thickness are increased by the same multiple, an increase in blockage area results in a higher temperature peak. Increasing the area of blockage necessitates a longer duration for both the temperature and the velocity fields to revert to equilibrium.

Prediction of Heat Transfer Coefficient of Two Phase Flow Sodium Reactor Based on Artificial Neural Network

Liquid sodium metal is one of the main coolants for fast neutron reaction in nuclear reactors with high heat load and high thermal conductivity. Sodium reactor is one of the most promising and commercialized reactor types in modern nuclear power system. Because the chemical properties of liquid sodium metal are very active, so in a sealed environment, it is easy to react with oxygen, water and other substances, resulting in an explosion. Therefore, it is necessary to design a reasonable experimental circuit and conditions under the appropriate experimental environment to carry out the experiment. With liquid metal sodium as the experimental medium, this paper studies the flow heat transfer characteristics of a two-phase flow sodium reactor by establishing a mathematical model. The key heat transfer parameters in a two-phase flow sodium reactor are predicted by building an artificial neural network model, and the parameters affecting the network performance are adjusted by the artificial neural network. The error of the final result predicted by the adjusted artificial neural network model is within the range of ±10%. The prediction results show that the prediction by the artificial neural network model has higher prediction accuracy.

Towards atomistic modelling of solid Pb-O formation and dissolution in liquid lead coolant: Interatomic potential development

Microscopic description of solid Pb-O formation and dissolution in liquid lead coolant is important for the modelling of fast neutron reactors. For this purpose, in this work we develop interatomic potential models for lead melt with dissolved oxygen. Potentials fitting is based on ab initio density functional theory (DFT) calculations and quantum molecular dynamics. Two different potential models are trained on the DFT data using the force-matching method: a computationally cheap embedded atom method (EAM) potential and a computationally expensive machine-learned moment tensor potential (MTP) that allows reproducing DFT data very accurately. The potentials are further validated to reproduce the properties of lead melt and the coordination of oxygen atoms in molten lead. The diffusion coefficient for oxygen in molten lead is calculated and compared with the available experimental data. We analyse how these interatomic potential models describe the crystal structure of the solid lead oxide to consider PbO nucleation and dissolution in molten lead. We extensively discuss the arising problems with the description of long-range electrostatic interactions. The possibility of matching the experimental data on the oxygen solubility limit by tuning the EAM potential is shown.

Evaluation of Reaction Rate Distribution Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, reaction rate distributions of the fission reaction rates of 239Pu, 235U, and 238U and the capture reaction rate of 238U were measured using activation foils during its system startup test. The measurements in the core and radial blanket regions were evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies were examined through a comparison with calculations. The reaction rate data measured in Monju were confirmed all reliable and useful as the validation data. The fission reactions of 239Pu, 235U, and 238U can be validated with an accuracy of a few percent in the core and blanket regions. The capture reaction of 238U in the core region also can be validated with a similar accuracy, whereas a precise calculation of the foil cross section is necessary to consider the resonance shielding effects of the surrounding fuel pins and a foil.

Applicability to DEMO breeding blankets of neutron measurements techniques from fission reactors

Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy.The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors.The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding.

Assessment of radiation environmental risks for fish fauna of the cooling pools of a number of operating nuclear power plants

Relevance. The need to assess radiation risks for the main types of fish in pools that are under the constant influence of the nuclear power plants. Aim. To assess the radiation environmental risks for the ichthyofauna of cooling pools of the operating nuclear power plants, to analyze the sources and features of formation of the radiation risks at different stages of their operation. Objects: monitoring studies of radiation environmental risks for reference fish species in cooling reservoirs of the Leningrad, Beloyarsk and Novovoronezh NPP. Methods. Calculation of the exposure dose rate and quantitative assessment of the risk of exposure to technogenic radionuclides of the ichthyofauna of the cooling pools of the Beloyarskaya, Leningradskaya and Novovoronezhskaya NPPs were carried out using the international computer complex of the updated version of ERICA Tool 2.0. Results. The paper demonstrates that the radiation ecological risks for the ichthyofauna in the cooling pools of the operating nuclear power plants are caused by the combined effect of radioisotopes of the different origin. The features of the formation of the radiation risks for each pool are specific and depend on the current radiation situation. The cooling pool of the Beloyarskaya NPP is characterized by a pronounced dependence of the radiation load on the types of operating power units. The ichthyofauna experienced the greatest radiation risks during the operation of the first two power units with thermal reactors AMB-100 and AMB-200. The decommissioning of the first power units and the construction of the units with the fast neutron reactors (BN-600 and BN-800) led to a significant reduction in the radiation load on the ichthyofauna of the cooling pool. At present, the radiation risks are caused mainly by the influence of 90Sr, which is almost not discharged from the Beloyarskaya NPP into the cooling pools, but has a predominantly global origin. Therefore, the formation of the risks is more associated with atmospheric fallout of 90Sr, and not with the operation of the nuclear power plant. The radiation risks for the ichthyofauna of the Leningradskaya NPP cooling pool are caused by the combined effect of 137Cs as a result of the radiation accident at the Chernobylskaya NPP and radionuclides with induced activity (14C; 60Co; 3H 54Mn and 65Zn), the origin of which may be associated with the operation of the nuclear reactors of the NPP. The level of radiation risk for the fish fauna of the Novovoronezhskaya NPP cooling pool for 2009–2019 increased 2.2 times due to an increase in the content of 60Co in many components of the pool. As a result, 81.7% of the radiation risks in demersal fish species were formed by the influence of 60Co. It cannot be ruled out that the increase in the 60Co content in the cooling pool of the Novovoronezhskaya NPP is a consequence of radioactive contamination of the groundwater as a result of the 1985 accident. The ecological risks for the fish fauna in all cooling pools are two to three orders of magnitude lower than the maximum allowable hazard coefficient. Thus, the normal operation of the nuclear power plants does not create unacceptable radiation risks that pose a real environmental threat to the fish fauna of the cooling pools.

TRU utilization and MA transmutation in thorium-based fluorinated molten salt fast reactor

In this paper, a thorium-based fluorinated molten salt fast neutron reactor is taken as the research object, and the utilization of thorium resources, reuse of transuranic waste (TRU), and transmutation of minor actinide (MA) in fast reactor are the research purpose. The fast reactor adopts a multi-layer core structure, in which the fuel is divided into an inner fission fuel zone and an outer breeding fuel layer, respectively loaded with fluorinated molten salts LiF+ThF4+XF4 (where X is 233U, TRU, or 233U+MA) and LiF+ThF4. In addition to serving as nuclear fuel, these two molten salts also act as coolants in the primary circuits.Through the cross sections of nuclear reactions, transmutation chain of actinide nuclides, and the particle population density formula of actinide nuclides in nuclear reaction equilibrium, the possible nuclear reactions of actinide nuclides in thorium-based fast reactors are analyzed. The calculated results show that the thorium-based fast reactor has the following characteristics: fast neutron energy spectrum, negative temperature reactivity coefficient, effective utilization of thorium resources, the ability to breed fissle fuel such as 233U, production of high-purity radioisotope 238Pu material, and the ability to incinerate Pu and MAs in the TRU spent fuel.In the 233U fuel scheme, the 233U fuel is breeded, breeding ratio of fissile fuels is 1.21, the reaction products contain very little Np, and almost no Pu, Am, and Cm long-lived radionuclides. In the TRU fuel scheme, the annual capacity to incinerate MAs and Pu is 1072.6 kg, while producing a large amount of 233U and 233Pa fuel (together 1162.5 kg/year).In the 233U+2.0 wt% TRU-MAs fuel scheme, 127.35 kg of MA waste can be incinerated annually, and their transmutation rate is 15.95%, which is equivalent to the annual production of MA nuclear waste from 4.9 conventional 1000 MWe light water reactors. Here, TRU-MAs is the mixtured MA fuels after separation of Pu from the TRU fuel. At the same time, radioactive fuel 238Pu, which can be used for space nuclear power, is produced with an annual yield of 87.89 kg and an isotopic abundance of 87.57%. If the added TRU-MAs fuel is replaced with 2.0 wt% 237Np fluorinated salt, the annual yield and isotope abundance of 238Pu are further increased to 130.24 kg and 95.4%.

Радиологическая безопасность населения и персонала при двухкомпонентной ядерной энергетике

The Fundamentals of State Policy in the Field of Nuclear and Radiation Safety of the Russian Federation, approved by the President of the Russian Federation, indicate that the goal of state policy in this area is to ensure the protection of the population, considering modern requirements. The IAEA Fundamental Safety Principles emphasize the need to protect both current and future generations from radiation risks. The achievement of these safety goals is ensured by the development in Russia of new nuclear energy system based on a closed nuclear fuel cycle (CNFC) and fast neutron reactors (FNR). One of the objects of the new nuclear energy system is the Experimental Demonstration Energy Complex (EDEC) consisting of the BREST-OD-300 FNR, a reprocessing module and a fuel fabrication/refabrication module, located on the territory of JSC «SHK» in the Tomsk region. This article provides assessments of radiation carcinogenic risks for the population living in the 30-km zone of JSC «SHK» under normal operating conditions and for workers after potential emergency situations at EDEC facilities. The conditions for ensuring the protection of future generations of people are also justified. Radiation risk assessments are obtained directly from the dynamics of equivalent doses in human organs and tissues and modern risk models recommended by ICRP Publication 103. The predicted values of the risks of radiation carcinogenesis for the population and personnel are significantly lower than the current limits of NRB-99/2009 for both normal operation and potential exposure during accidents. It has been shown that transmutation of minor actinides during the development of CNFC based on FNR ensures the safety of future generations of people: it reduces the carcinogenic risk of americium by 213 times, neptunium by 101 times, and curium by 47 times. The effect of radiological equivalence of radioactive waste and natural uranium raw materials is achieved after 99 years of radioactive waste storage. Thus, the priority direction for the development of new nuclear power is the energy sector of the CNFC based on the FNR.

Радиоэкологическая оценка реакторных установок на основе сценариев проектных и запроектных аварий

The methodical approach based on scenarios of design and beyond design basis accidents was developed for comparative radioecological assessment of reactor facilities. The dynamics of possible radiation doses to the thyroid and the whole body of the public resided nearby the Russian nuclear power plants equipped with the VVER-440, VVER-1000, VVER-1200 and BN-800 reactors was estimated. With the use of the conservative approach for each accident scenario, as well as with the use of acute and medium-term dose criteria, radiation exposure indices were calculated for each accident scenario. Reactor facilities are ranked according to an index summarizing the accident scenarios features and the probability of their realization, the height of radioactive emissions and the activity of dosage-forming radionuclides. The potential radioecological danger of Russian reactors to the population has been reduced. The lowest estimates of the summarized emergency risk are typical for the advanced VVER-1200 reactor of the 3+ generation and the fast neutron reactor BN-800. The most severe emergency scenarios for each considered reactor are highlighted. The VVER-440 reactor has the highest number of severe projected accidents exceeding the dosage criteria envisaging evacuation of the population. The directions of further research related to the analysis of the environmental radiological effects, the justification of radiation protection measures and the development of operational intervention levels in the acute post-accident period are outlined.

Generating multi-group cross-sections using continuous-energy Monte Carlo method for fast reactor analysis

The deterministic two-step method, comprising multigroup cross-section generation and core calculation, is widely applied in fast reactor design and analysis. Monte Carlo (MC) methods with continuous energy and fine geometry provide high-precision cross-sections essential for advanced fast reactor neutronics analysis. This paper presents an analysis of integrating MC-generated homogenized cross-sections with various core solvers, demonstrating their effectiveness and potential improvements in fast reactor simulations. For diffusion core calculations, the superhomogénéisation (SPH) technique reduces control rod worth overestimation from 13.5% to 0.35% in the MET-1000 benchmark, improving power distribution predictions. In transport core calculations, the flux-moment homogenization technique (MHT) addresses reactivity overestimation by incorporating cross-section anisotropy, reducing error by 698 pcm. For Method of Characteristics (MOC) core calculations, transport-corrected multigroup cross-sections yield high precision in pin-by-pin power distribution for a 100 MWe lead-based fast reactor benchmark. While MC methods require significant computational resources, such as 62 CPU-hours for the MET-1000 core and 85.5 CPU-hours for the 100 MWe lead-based fast reactor core, they offer flexibility in geometry modeling. This work highlights MC multigroup cross-section generation methods applicable to diffusion, MOC, and transport core calculations for fast reactor analysis, suggesting further exploration into their performance in various reactor parameters and computational efficiency.

In-situ study on factors impeding dislocation motion in Fe9Cr1.5W0.4Si F/M steel during high temperature heating

Ferritic/martensitic (F/M) steels are promised candidate materials for the lead-cooled fast neutron reactor. However, their poor high temperature performance limits the highest in-core temperature, decreasing the reactor economics. High dislocation mobility directly related to the low strength of F/M steels but the underlying mechanisms remain unexplored. Here, in-situ heating was performed on irradiated Fe9Cr1.5W0.4Si ferritic martensitic steel in the temperature range of 873–923 K, and the interaction between dislocation lines and obstacles was investigated to reveal the mechanism that led to easier dislocation gliding. Dislocations tended to glide in several planes at the same time and only glided as a whole in one plane occasionally. It was discovered that dislocations usually glided in high index planes such as {123} planes. When moving dislocations intersected with forest dislocations, a new mechanism was found where the segments for different dislocations with the same Burgers vector could exchange, which was important for the moving dislocation to cut through the forest dislocations without decreasing their mobility. The reactions between dislocations and <100> loops included absorption, break away and rebound. Meanwhile, statistical results indicated that the highest ratio was about 48% for the case where <100> loops were restored and dislocations broke away, demonstrating that the <100> loops were not strong obstacles when interacting with moving dislocations at high temperature.

The Contributions to Registration Efficiency of The Fast Neutron Reactions on The Nuclei of The Heavy Oxide Scintillators

The results of the study of the contributions of the interaction reactions of fast neutron sources of 239Pu-Be and 252Cf to the counting efficiency of registration by oxide scintillators CdWO4, ZnWO4, Bi4Ge3O12 and Gd2SiO5, presented. The amount of gamma quanta per input neutron emitted from final nuclei excited in the reactions of inelastic scattering (n, nʹγ)in, resonant scattering (n, n)res and capture (n, γ)res and radiation capture (n, γ)cap was measured. PMT R1307 operating in single-electron mode was used as a photodetector, the background rate was ~ 5*103 s-1. The measured efficiency ε for scintillators ø40x40 mm was 752 for ZWO, 532 for CWO, 37 for GSO, and 23 for BGO in "counts/neutron" units, measurement error rate ~ 3-5%. The formation of the detector response is influenced by the parameters of the scintillator nuclei, such as the values of the interaction cross sections in the resonance region, the density of nuclear levels of the final nuclei, the lifetime of excited nuclear states, the upper limit of the resonance region of the cross section, as well as the scintillation time and geometric parameters of the scintillators. A phenomenological model of the response of an oxide scintillator to fast neutrons is proposed.

Adaptation of the turbulence model to the heat transfer in a rod bundle

Fast neutron reactors using liquid metal or gas coolants are one of the most actively studied fields of research. Increased safety requirements and insufficient knowledge of the thermohydraulic characteristics of these coolants require validation of RANS simulations via information from experiments or DNS. This information can also be applied to improve the RANS turbulence model ifnecessary.In this paper, a series of thermohydraulic DNS simulations of the uniformly heated rod bundle with a relative pitch of 1.4 have been performed to improve RANS. A total of 12 test cases with four different Prandtl numbers of Pr = 0.0043, 0.0154, 0.0324 and 0.71, which correspond to liquid metals and gas, and three Reynolds numbers of Re = 2500, 5000 and 10,000 are considered. The present simulation results show good agreement with the experimental and DNS data. It is revealed that the intermittent flow regime forms at Re = 2500. Furthermore, the turbulent Prandtl number is structured with a maximum of up to y+≈25 and, from y+≈50 onwards, in most cases takes on a constant value. By increasing the Prandtl or Reynolds number, the average value of the turbulent Prandtl number tends to be 1.0. It was also revealed that the turbulent contribution to heat transfer is linear to the Reynolds number for the selected coolant. A new formula for the temperature profile was also obtained, which shows better agreement with DNS data compared to its analogues.As a result, the selected RANS model has been improved (the maximum integral error dropped from 49% to 4%) by the new model of the turbulent Prandtl number and by introducing model coefficients from DNS.

The effect of static stresses on the acceleration of the decomposition processes in the ferritic-martensitic steels in conditions of neutron irradiation to a dose of 85 dpa

Behavior of the corrosion-resistant EP823 steel (Fe-11Cr-Mn-Mo-Nb-V-W-Ni-Si-0.17C) and its ODS modification, hardened by yttrium, yttrium-titanium oxides (EP823-ODS) in conditions of static mechanical stress under high dose exposure in a fast-neutron reactor was studied using Mössbauer spectroscopy and electron microscopy. Mechanical stresses were revealed to result in a deformation of the fuel elements walls and the acceleration of the decomposition processes in steel structure with a release of chromium into the intermetallic χ phase and to the Me23C6 carbide formation. It was also shown that the key factor accelerating the structure decomposition upon neutron irradiation to a dose of 85 dpa at 505 °C is anisotropy of stresses between longitudinal and transverse directions of fuel cladding tube surface.

Improvement of LLFPs transmutation efficiency by moderator loading in lead-based fast neutron reactor

The long-lived fission products (LLFPs) are one of the major impediments to the development of nuclear energy due to their longevity and toxicity. The fast neutron reactor, with its high neutron flux and excessed neutron, is a strong candidate for transmuting LLFPs. In this investigation, 79Se, 99Tc and 129I were selected as the representative nuclide of LLFPs, and then neutron moderators were loaded into the LLFPs assembly with LLFPs to evaluate the effect of different loading approaches on transmutation efficiency by the transmutation performance of LLFPs. The results show that the optimal loading strategy of moderator could increase the TR and SR with 2.35 (1.01) and 284.80 (151.11), 0.36 (0.28) and 8.06 (7.88), 1.69 (0.69) and 11.45 (7.80), respectively. It is suggested that it can transmute an additional 4500 g, 1000 g and 6600 g of these three LLFPs per year.

Investigation of fast neutron reactions in natural lutetium

The cross-sections and isomeric ratios of the cross-sections for the reaction 176Lu(n,γ)177m,gLu were obtained for the first time on a base of experimental and evaluated data with the MCNP simulation of scattered neutron spectra. The cross-sections of the reactions 175Lu(n,p)175Yb, 175Lu(n,α)172Tm, and 176Lu(n,α)173Tm were precised at a neutron energy of 14.6±0.3 MeV and used for checking up our cross section determination procedure. The account of scattered neutrons to the nuclear reaction was estimated with the MCNP 4c code and TALYS-1.96, EMPIRE 3.2.3 codes calculations. Even with a proper consideration of scattered neutrom spectra, the 176Lu(n,γ)177m,gLu reaction cross section appeared rather greater than the calculations with the TALYS, EMPIRE codes and evaluations within the ENDF/B-VIII.0, TENLD-2021, JEFF-3 data bases and libraries.

Study of Uranium and Thorium Fuels in Breed-and-Burn Mode

The formation of fissile nuclei through breeding conversion is a hot topic in academic research, as it provides a continuous source of nuclear fuel for nuclear reactors. Fast neutron reactors, which have been extensively studied, use natural uranium or low-enriched uranium as the nuclear fuel, achieving burning after uranium-plutonium conversion. Thorium, as another potential fissile fuel, can theoretically be converted into nuclear reactor fuel through the thorium-uranium cycle. In this study, the physical evolution process of nuclear fuel in a specific core parameter is simulated using the Monte Carlo program, and the performance differences of the thorium-uranium cycle and uranium-plutonium cycle in achieving in situ breeding-burning (Breed-and-Burn, B&amp;B) mode are analyzed using neutron balance analysis method. The optimal ratio conditions for achieving self-sustained B&amp;B burning in a thorium-uranium fuel mixed core are investigated. The study shows that for traditional solid-state nuclear reactor cores with a lower fuel proportion, thorium-breeding fuel has poor neutron economy compared to uranium-based breeding fuel, making it more difficult to achieve the B&amp;B mode.