Space Nuclear Power Systems Research Articles

Simulation and Mitigation of False Data Injection Cyberattacks on a Space Nuclear Reactor Power System

This paper investigates the response of the DynMo-CBC space nuclear reactor power system to simulated cybersecurity attacks during a startup transient and demonstrates the effectiveness of the mitigation measures. The system nominally generates 134 kW(electric) continuously for 12 years and does not have a single-point failure in reactor cooling and energy conversion. The reactor core is divided into three hydraulically independent sectors, each having a separate loop with a single shaft, closed Brayton cycle (CBC) turbomachinery unit. A He-Xe gas mixture with a molecular weight of 40 g/mol cools the reactor core sectors and is the CBC unit’s working fluid. This paper examines the effects of simulated false data injection attacks (FDIAs) on the operation parameters of the power system. The simulated FDIAs decrease or increase the external reactivity insertion beyond nominal to cause spikes in the reactor’s power and temperatures. The results demonstrate the effectiveness of the programmable logic controller regulating the control drums’ drive motors. It mitigates the effects of the simulated FDIAs on the transient operation of the power system and shortens the recovery time after the termination of the simulated cyberattacks.

Design and optimization of space nuclear power system with sCO2 Brayton cycle on Mars

As human exploration of outer space deepens, the demand for space power increases. This requires not only higher thermoelectric conversion efficiency but also a more compact system. Space nuclear power based on the sCO2 Brayton cycle is better suited for the surface of Mars due to its high efficiency and low ratio of mass to electric power. This paper designs a 6 MWt sCO2 Brayton cycle power generation system, and the following system configurations are mainly studied including Simple Heat Recovery Cycle (SRC), Re-Compress Heat Recovery Cycle (RC), Reheating Heat Recovery Cycle (RRC), and Reheating Re-Compress Cycle (RHRC). A thermodynamic calculation model of the system was developed based on the aforementioned cycle configurations. Additionally, a quality assessment model was developed separately for each component within the system, according to its respective category. The effects of parameters such as temperature, compressor pressure ratio, and bypass ratio on the system efficiency and total system mass are also examined for each of the four-cycle configurations. The study conducted a global multiparameter optimization using a genetic algorithm to optimize system efficiency and the ratio of mass to electric power. The RHRC configuration achieved the highest cycle efficiency at 39.47%, while the SRC cycle had the lowest system efficiency at 35.46%. The RRC configuration had the minimum ratio of mass to electric power at 7.419 kg/kWe with an efficiency of 33.13%.

Numerical study of anisotropic scattering and solar radiation impact on liquid droplet radiator performance

The Liquid Droplet Radiator (LDR), a lightweight frameless radiant heat rejection system, is widely regarded as the optimal heat rejection solution for megawatt-scale space nuclear power systems. The LDR dissipates heat to space through its droplet layer, which plays a crucial role in determining its radiative rejection capacity. Radiation beams undergo a series of emission, scattering, and absorption processes within the layer before being emitted into space. Prior studies typically assume isotropic scattering in this process. However, this study develops a three-dimensional model for the LDR’s droplet layer based on Mie theory and Monte-Carlo methods to investigate the impact of anisotropic scattering and solar radiation. Our calculations reveal that the working fluid exhibits strong forward anisotropic scattering characteristics. Under the anisotropic assumption, the heat rejection capability of the LDR is 27–33 % higher compared to the isotropic assumption. Regarding the temperature distribution within the LDR, under the anisotropic assumption, the layer’s temperature is 6 K lower at the XY boundary surface and 3 K higher at the layer median, as opposed to the isotropic assumption. We introduce the concept of the probability of a light beam escape to elucidate the distinct beam transmission mechanisms and temperature distribution. When the distance between the LDR and the sun is in the vicinity of 1 AU, the impact of solar radiation on the LDR’s ability to operate is on the scale of 1E-5, which is almost negligible. This study lays down a theoretical groundwork for calculating the heat rejection capacity during the LDR design phase.

Efficiency and mass optimization for space nuclear power systems with closed Brayton cycle loops

Space nuclear power systems should be strictly limited in mass and size. Lack of detailed designs, mass estimation in the general design phase usually relies on simple and empirical models that leave too much design margin from the limits. Accurate mass estimation of space nuclear power systems is key to minimize the design margin and maximize the system performance in the general design phase. In this paper, new mass estimation models are proposed for the main components of space nuclear power systems, including nuclear reactor, shield, turbine, compressor, heat exchanger, and thermal radiator. The mass estimation models are more accurate by incorporating principles of nuclear physics, thermal hydraulics, and structural mechanics to the component designs. Combined with thermodynamic models and genetic algorithm, the MEGREZ code is developed to optimize the system performance by setting efficiency and mass as dual objectives. The code verification results indicate that the deviations between and the Jupiter Icy Moons Orbiter project parameters and the code calculated values are only 0.4% for system efficiency and 2.1% for system mass. The optimization for the Jupiter Icy Moons Orbiter project shows that the optimal molecular weight of helium-xenon mixture is the minimum value for single-stage turbomachinery limits. Additionally, the impacts of potential technological degradation, such as increased heat exchanger thermal resistances, higher frictional factors, and decreased turbomachinery efficiency, on system performances are discussed. The results show that turbomachinery efficiency presents the strongest influence on the system-specific mass; a 0.04 reduction in the polytropic efficiency of the compressor or turbine results in more than a 10% increase in system-specific mass.

Shielding mass estimation model for gas-cooled space nuclear reactors

The shielding is expected to be “mass-effective” for space nuclear power systems (SNPs) because it accounts for a significant portion of the total weight. Hence, the shield mass should be estimated rapidly but accurately especially in the early stage of SNPs’ design. In this paper, a simple mass estimation model was developed based on the preliminarily optimized shield design for the Jupiter Icy Moons Orbiter (JIMO) reactor. In the model, the reactor-induced 1 MeV equivalent silicon neutron fluence and photon absorbed dose at the unshielded payload were predicted by a circular surface source hypothesis at the core bottom, and the proportional coefficients were determined by the Monte Carlo calculation. Moreover, the attenuation factors of the shield were derived by solving the neutron/gamma coupled transport process within the shield body, and the correlations were established between the attenuation effect and the shield thicknesses. Afterwards, when given the radiation limits at the payload, the shield geometry and associated mass could be easily determined. Several detailed calculations corresponding to the variation of shield design were used to verify the estimation model. Those variations include the thickness of the Be-B4C layers, the effective density of the B4C and the number of the shielding layers. Additionally, the potential changes of the reactor such as the neutron energy spectrum and the structural materials in the core were also considered to evaluate the applicability of the model. Totally, the estimation model agrees with the detailed calculation within 15 % in a wide range, which suggests the model can be incorporated into the system code to help optimize the overall performance of the SNPs.

Thermodynamic performance of an innovative space nuclear Brayton cycle with S–N2O

The supercritical Brayton cycle system is renowned for its characteristics of elevated high-power density, compact structure, lightweight design, and extended operational lifespan, rendering it as a preeminent energy conversion approach for space nuclear power systems. Using nitrous oxide (N2O) as the working fluid in a supercritical Brayton cycle system enables heightened efficiencies at comparatively lower cycle temperatures. This study examined four distinct layout configurations, culminating in an in-depth analysis of their respective thermodynamic performances. Among the configurations analyzed, the newly proposed recompression Brayton cycle with partial cooling and heat recovery (RBC-PCHR) layout exhibited the best thermodynamic performance compared to other cycle configurations. This superiority can be principally attributed to augmented heat recovery procedures and reduced compression workloads. A detailed multi-objective optimization strategy is subsequently embarked upon, building upon the insights gained from the sensitivity analysis pertaining to the key operating parameters. The optimization outcomes show significant enhancements, with an increase of 1.37% in ηth and 32.18% in MBRU, along with a marginal decrease of 0.70% in ηex.

Conceptual design for a 5 kWe space nuclear reactor power system

Enhancing the capabilities of unmanned space exploration, such as satellite monitoring and space science missions, requires efficient and reliable nuclear power systems. A viable solution is found in the 1–10 kWe power level of space nuclear reactor power systems, offering advantages such as a manageable research and development process, and relatively low investment requirements. This paper introduces a conceptual design for a 5 kWe space nuclear reactor power system, outlining its components and characteristics. The study includes a thorough analysis of potential challenges, encompassing heat pipe failure accidents, re-entry scenarios, and weight estimation considerations. The results demonstrate that the proposed space nuclear reactor power system effectively meets the safety requirements. The total mass of the power system is estimated at approximately 1.5 tons, with a specific mass of around 300 kg/kWe. This research contributes valuable insights for the design of space nuclear reactor power systems operating within a similar power range.

Performance analysis of indirect contact heat pipe radiator for space nuclear power system

The heat pipe radiator, with excellent thermal conductivity and anti-collision performance, is expected very suitable for heat emission in nuclear-power spacecraft. Indirect contact heat pipe radiator could operate longer life with lower capital cost. In this paper, the fluid-solid coupling heat transfer method is integrated to solve the energy model and S–S radiation model based on the k-epsilon model. Optimization objectives, heat dissipation performance and temperature uniformity are adopted to optimize the radiator structure. A comparison analysis between the ordinary and optimized radiator is conducted based on energy and entropy variation. The results show that the temperature uniformity and heat dissipation characteristic of the radiator after the optimization of connection mode of heat pipe and coolant pipe, fin shape, and fins welding direction have been improved without changing the area of fins and the weight of radiator. Through the comparative of radiators after changing the working conditions, some useful conclusions are drawn, which provide the theoretical foundation for the experimental research and engineering practice.

Optimization of thermodynamic performance and mass evaluation for MW-class space nuclear reactor coupled with noble gas binary mixtures Brayton cycle

Space exploration technology is an important indicator of society's technological level. A space reactor coupled with a Brayton cycle is preferable for megawatt-scale space power systems. Noble gas binary mixtures have high chemical stability, heat transfer performance, and compressibility, making them the principal choice of working fluid for the space reactor Brayton cycle, which is also the key factor affecting the thermodynamic performance and mass of the system. This study developed thermodynamic performance and mass evaluation models for the space nuclear Brayton cycle and discovered the inherent relationship between system thermodynamic performance and mass. The effects of noble gas binary mixtures on system performance and mass were investigated. The results indicated that the elevated molar mass of noble gas binary mixtures reduced the aerodynamic load and mass of the turbomachines and increased the mass of the recuperator. There are optimal values of the total mass, specific mass of the system, and working fluid composition. Helium-xenon mixture is the optimal working fluid because it can achieve the highest thermodynamic efficiency and lowest mass. Furthermore, the optimal scheme of the helium-xenon Brayton cycle for a space nuclear power system was obtained by multi-objective optimization. Its power generation efficiency, specific mass, and helium molar fraction in the helium-xenon working fluid are 29.04%, 5.65 t∙MW−1 and 77.5%, respectively. This study provides a reference for the design and optimization of space nuclear power systems.

Numerical investigation of nozzle length effects on Rayleigh jet breakup behaviors and droplet formation properties at different modulation amplitudes in liquid droplet radiator

Uniform droplet stream generation is required for a liquid droplet radiator to achieve the efficient heat dissipation of space nuclear power systems. In this study, Rayleigh jets breaking into continuous droplet streams at different modulation amplitudes are numerically simulated to investigate the nozzle length effects on jet breakup behaviors and droplet formation properties. First, simulated results confirm that the growth of secondary instability wave on jet free-surface causes the jet breakup pattern transition from downstream pinching to upstream pinching. As the modulation amplitude is from 0.04 up to 0.16, the nozzle length range where the upstream pinching occurs in the jet breakup is broadened from null to 1–5. The essential reason for the increase of jet breakup length is the reduction of fundamental amplitude and the increase of full velocity relaxation length. Then, a comprehensive prediction model for the velocity-modulated jet breakup length is developed in agreement with the experiments, and the average relative error is below 7.38 %. Finally, two types of satellite droplet formation regimes including rear-merging and forward-merging satellite droplets are identified by the oscillation mode of the jet tip. It is found that the primary droplet generation frequency is equal to the external disturbance frequency and independent of nozzle length and modulation amplitude. Increasing the nozzle length and modulation amplitude can both increase the satellite droplet diameter and primary droplet spacing, while the primary droplet diameter is almost constant at around 1.85.

Study on load tracking characteristics of closed Brayton conversion liquid metal cooled space nuclear power system

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop's helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system's main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.

Preliminary analysis of maximum hypothetical accident for a solid core space nuclear reactor power system

In order to study the safety characteristics of the solid core space nuclear reactor power (SNRP) system under the maximum hypothetical accident, a two-dimensional entire core transient heat transfer analysis model was established, and the key parameters response characteristics of the ultra-small lithium-cooled SNRP under two maximum hypothetical accidents, namely, loss of heat sink (LOHS) and loss of coolant accident (LOCA), were calculated and analyzed. In the maximum hypothetical accidents, the coolant cooling capacity is lost, thus the core decay power is discharged into space only through the radiation of the residual heat removal system and the side surface of the reactor vessel. During the accidents, the heat transfer of the core deteriorated, and the fuel temperature may rise to the melting point, resulting in radioactive leakage. The results show that: (1) In the LOHS accident, the maximum fuel temperature reaches 2150 K at 550 s, and the pressure of the primary system volume accumulator continues to increase to the set system pressure safety limit of 2 MPa, resulting the primary loop overpressure failure. And the fuel matrix temperature is close to the set cladding limit temperature of 2200 K; (2) In the LOCA, the deterioration of heat transfer in the core makes the temperature increase rapidly, reaching a maximum of 3016 K at about 630 s, which is very close to the melting temperature of UN fuel 3123 K. As the decay power decreases, the maximum core temperature decreases to less than 1600 K after 24 h of the accident. The auxiliary cooling system of the solid core SNRP system under the maximum hypothetical accident is optimized, and the design parameters of the auxiliary cooling system meeting the safety requirements are obtained.

Research and analysis of the thermal and control characteristics of the plate-type fuel assembly for the supercritical CO2 reactor

Small nuclear reactors can be applied to micro grid power generation, island and space nuclear power systems. The plate-type fuel assembly has a compact design structure, low fuel core temperature, large heat exchange area, and high heat exchange efficiency, which is widely used in small nuclear reactors. To supplement the lack of research in S-CO2 small nuclear reactor, The Brayton cycle reactor system analysis program of S-CO2 plate-type fuel assemblies (BRESA-PFA) is developed, and the reactor control system program is developed to simulate the transient condition in this paper. This article establishes a fuel assembly flow and heat transfer model and a control rod control model using Modelica language, achieving physical and thermal coupling and power control of the reactor. The steady-state operating parameters of the developed program BRESA-PFA were studied, and the flow distribution, coolant and fuel temperature distribution of BRESA-PFA were obtained. The flow distribution conforms to the experimental parameters of CARR, and the fuel assembly radial temperature distribution is high at the center and low at the edge. After verification, the maximum fuel temperature did not exceed the limit value and met the design requirements. Transient characteristic analysis was conducted to study the rod lifting strategy during the startup process, the start-up and rod lifting process of BRESA-PFA is N2-N1-G2-G1, using the logic of step control. Research on reactor control system response under loss of coolant accident, after the accident, the coolant flow suddenly decreased to 65 % of the rated value, and the reactor power also decreased to 65 % FP, G2 and G1 control rods were inserted downwards, the coolant temperature fluctuates by 1 %∼2%. The reactor control system was able to respond quickly, ensuring the stability of coolant temperature, proving the safety and reliability of BRESA-PFA, and providing theoretical reference and scheme support for the research of S-CO2 small nuclear reactors.

Mass optimization of a recompression supercritical nitrous oxide and helium power system for space exploration

Due to high output power, long lifetime, high efficiency, and compact structure, the space nuclear power (SNP) system is considered an ideal choice for future large-scale and deep space missions. The system's mass, volume, and thermoelectric conversion efficiency directly affect and determine the system's performance compared with ground nuclear devices. The mass of the space nuclear power system is a vital parameter due to the limitation of cost, volume, and transportation. This study designed and optimized the recompression supercritical with nitrous oxide and helium (N2O–He) Brayton cycle with a small modular reactor for mass. Each Brayton cycle component model was developed to estimate the cycle performance systematically. The mass estimation model of the Brayton system includes a nuclear reactor, radiation shadow shield, Brayton cycle unit, recuperator, and radiator. A small modular reactor was designed to predict the minimum mass reactor suited to the given cycle conditions. The system mass optimization explored the trade-offs between the reactor, radiation shield, Brayton cycle unit, recuperator, and radiator to obtain the minimum mass of the space nuclear energy system to satisfy the operating requirements. Sensitivity parametric investigations are considered to evaluate the effects of crucial decision variables on the whole Brayton cycle mass. Furthermore, multi-objective optimization is performed to find optimum operating parameters to minimize the system mass. The results illustrate that the specific mass of the system is less than 20 kg/kW. The total mass of the whole system is 5605.93 kg using near-term materials, of which the Baryton Rotating Unit, radiator, and shadow shield mass dominate the total mass. The Brayton Rotating Unit, radiator, and radiation shield account for 44.64%, 32.53%, and 14.29%, respectively. It is concluded that after the dual-objective optimization analysis, the mass of the cycle decreased by 18.44%, which used the existing technical materials.

Study on the load loss characteristics of a space nuclear power system with multi Brayton loops

Space nuclear power system (SNPS) is very attractive for the future civil and military space demands. SNPS with multi Brayton loops (SNPS-MBL) can achieve high-power output for space applications. The accident characteristics is the key to safe and efficient operation of SNPS-MBL. In this study, the dynamic model of the SNPS-MBL was established, the characteristics of SNPS with dual Brayton loops (SNPS-DBL) after single or dual Brayton loop (BL) load loss were studied. The results indicated that load loss was a serious accident. The positive torque on the turbine shaft after load loss increased the rotational speed rapidly to maximum value at 69.15 krpm exceeded the upper limit. Countermeasures should be taken after load loss. There was a coupling effect between the two BLs after load loss of single BL. load loss of the accident BL increased speed rapidly to 69.15 krpm, and decreased the BL power generation to 0 kW. The increase of inventory in normal operation BL increased power generation. Total power generation of SNPS-DBL decreased from rated value 150.8–81.8 kW. Furthermore, the influence of BL number on the safety of SNPS-MBL under accidents was analyzed. The results showed that increase of BL number improved the safety of SNPS-MBL in case of one BL load loss. The maximum speed of accident BL was not related to the number of BL. This study provides a reference for the design and operation of SNPS-MBL.

Dynamic performance of the combined stirling thermoelectric conversion technology for a lunar surface nuclear power system

The space nuclear reactor power system is one of the most potential energy and power supplies for future lunar science exploration. Unlike the devices on Earth, the devices operating on the lunar surface require a higher level of intelligent and automatic system control. Thus, it is necessary to establish a comprehensive and accurate dynamic system model to improve the system's automatic control. In the present study, a dynamic system model of a lunar surface nuclear power system combined with a Stirling cycle was proposed. The dynamic system model consists of a reactor core model, a heat rejection model, and a Stirling thermoelectric conversion model (while the reactor core is simplified as a point heat source). MATLAB and Simulink platforms are applied to implement the system model. To verify the system model and analyze the dynamic performance, simulations of the several reactor startup processes, different reactor thermal power (reactivity disturbances), and different lunar surface temperature conditions were carried out. The results show that the effects of thermal power variation on the dynamic system model have a nonlinear delay. Furthermore, the lunar surface temperature significantly impacts the dynamic system's operation and response. The present study provides a practical tool for the transient analysis of the lunar surface nuclear reactor power system and theoretical support for the design of lunar surface nuclear reactors.

Ridge regression and artificial neural network to predict the thermodynamic properties of alkali metal Rankine cycles for space nuclear power

The alkali metal Rankine cycle system has great advantages in space nuclear power system. The investigation of thermodynamic properties of alkali metals is fundamental to develop thermodynamic model of thermoelectric conversion system. This paper build models for predicting the thermodynamic properties of alkali metals (potassium, sodium, cesium, rubidium) by statistics and machine learning. Pearson correlation analysis is performed to explain the correlation between measurable variables (temperature, pressure, specific volume, specific heat capacity) and unmeasurable variables (enthalpy, entropy). Ridge regression and radial basis artificial neural network models are established for the enthalpy and entropy of alkali metals respectively. Five evaluation indicators: residual, SSE, MSE, ME and R2 are selected to evaluate the accuracy of different types of prediction models. The practicability and extensibility of the prediction model are proved by case analysis. The results show that the prediction model of enthalpy and entropy based on radial basis neural network has high accuracy, and R2 can reach above 0.9. The entropy prediction model can even be stable at 0.999 when applied to different operating conditions and alkali metals. The new method proposed in this work solves the difficulty of experimental measurement in high temperature section to some extent, and opens a new idea for studying the properties of working medium. At the same time, the validation of the extrapolation feasibility of this method is of great significance to improve the thermodynamic properties database of alkali metals in the whole temperature range.

Numerical Simulation and Experimental Study on Gas Flow in an Open Lattice Structure for an Advanced Space Nuclear Power System

Helium flow in the rod bundle channel with an open lattice structure is an important phenomenon for the advanced gas-cooled nuclear core design. In this study, thermal analysis with helium flow in various channel designs is conducted based on CFD methods to determine a dimension-optimized rod bundle channel. An experimental study then follows in order to pick up an appropriate gas flow model in further numerical simulation. Finally, helium flow in the bundle channels consisting of 217 rods is simulated using this chosen flow model, which shows to satisfy the requirements of basic thermal analysis of a newly designed gas-cooled reactor with an open lattice structure. Generally, this work will contribute to the design and analysis of the future advanced space nuclear power system.

Performance analysis and optimization of heat pipe-based radiator for space nuclear power system

The weight of radiator affects the launch weight of spacecraft and the feasibility of engineering development. Heat pipe radiator is widely used in space heat emission because of its strong resistance to single point failure and excellent heat transfer performance. This paper analyzed the sensitivity of fin width, inlet temperature, condensing section length and coolant mass flow of a direct contact heat pipe radiator are analyzed on four types of heat pipe radiators. The optimal design parameters of two inlet – two outlet radiator are obtained by double objective optimization, and the optimal design parameters are Lf = 0.0893 m, Tf11 = 697.8371 K, lhpc = 0.7736 m, qm = 8.9184 kg/s. ANSYS fluent 19.0, a CFD code is used to simulate the flow and the heat transfer characteristics of radiator with the best design parameters. Which indicates the intersection of coolant pipe and annular pipe of the radiator should be thickened, and the fin shape should be optimized along the direction of the fin isotherm.

Comparative study on sequential and simultaneous startup performance of space nuclear power system with multi brayton loops

The closed Brayton cycle is an ideal energy conversion technology for high-power space nuclear power system (SNPS) because of its small volume, light weight, and stable operation. To increase the redundancy of the system, a SNPS with multi Brayton loops (SNPS-MBL) is adopted. Startup performance is an essential part of the off-design performance of SNPS-MBL. In this study, the thermal-hydraulic model of the SNPS-MBL was established. Two startup schemes, simultaneous startup and sequential startup, for SNPS with dual Brayton loops (SNPS-DBL) were formulated and the performance of the two schemes were compared. The results indicated that motoring power of external power source (EPS) was required to drive the rotation of the shaft at the initial startup stage. And the simultaneous startup of SNPS-DBL was stable and less time-consuming, but the motoring power of EPS was high; the motoring power of sequential startup scheme was less than that of simultaneous startup, but the coupling effect between Brayton loops led to the fluctuation of gas inventory and power in Brayton loop (BL), which is easy to cause the instability of the system. Furthermore, the influence mechanism of space environment temperature on sequential and simultaneous startup schemes were analyzed. The results showed that the environment temperature impacted the coupling effects of multi BLs during sequential startup. Elevated environment temperature redistributed coolant inventories between BLs during sequential startup and changed power of EPS and startup time of the two startup schemes. This study provides a reference for the design and operation of SNPS-MBL.