Loss Of Sink Research Articles

CFD Simulations of a Loss of Heat Sink Experiment in the McMaster Nuclear Reactor

The McMaster nuclear reactor (MNR) is an important research facility that not only provides researchers with neutrons for fundamental science, but also supplies the medical industry with isotopes used for cancer treatment. To ensure the safety and performance of the MNR, modeling of the thermal hydraulics during nominal and accidental conditions is required. For such a task, system codes are customarily used. While system codes can assess the safety aspects of complex thermal–hydraulic systems, the question arises whether such systems can be modeled appropriately in a three-dimensional manner, such as computation fluid dynamics (CFD), while capturing the thermal mixing of the coolant. Modeling the thermal hydraulics of nuclear research reactors using CFD is relatively new. Therefore, validating such methods against experimental data is of utmost importance. The validation of CFD is the main focus of this work. More specifically, the influence of the loss of heat sink on the pool temperature is assessed using the CFD code Ansys CFX. The validation basis was provided by an experiment performed at the MNR in March 2023. In this experiment, the secondary heat removal system was intentionally shut down, and the pool temperature was measured in a few positions. The results obtained by modeling the loss of heat sink in the MNR using Ansys CFX agree well with the experiment.

MUPSA-based evaluation of portable equipment utilization for risk reduction in nuclear power plants: Framework and case study

Based on the lessons from the Fukushima accident, new strategies to mitigate the spread of accidents in nuclear power plants have been established that consider portable equipment. The diverse and flexible coping strategies (FLEX) and multi-barrier accident coping strategy (MACST) based on the utilization of portable equipment are being adopted in case of an extended loss of all AC power and loss of ultimate heat sink. In this context, the establishment of specific strategies for utilizing portable equipment has emerged as an important issue. This study proposes a multi-unit probabilistic safety assessment (MUPSA)-based evaluation framework consisting of several steps to quantitatively assess the reduction in risk from the use of portable equipment at a multi-unit site, which can provide a basis for establishing optimal portable equipment strategies. According to the framework, after first defining the scope of the evaluation, risk-significant types of portable equipment are selected and the scenarios are analyzed. Risk reduction is then evaluated using modified MUPSA models according to the deployment of the portable equipment. A case study was conducted for an internal loss of off-site power (LOOP) and seismic LOOP focusing on how to evaluate the effectiveness of portable equipment through the developed framework.

Thermal-hydraulic characteristics analysis of unprotected accident and protection control strategy for helium-xenon cooled reactor system

The small modular helium-xenon cooled reactor has the advantages of simple and compact system structure, short construction cycle, light weight and small volume, which is very suitable for energy supply in remote area and small nuclear propulsion device. The research on accident characteristics and protection control strategies is very important to ensure the safe and reliable operation of the reactor system and mitigate the consequences of the accident. Therefore, the safety characteristics analysis program of helium-xenon cooled reactor system is developed based on Modelica language, and the transient verification of typical working conditions is carried out. The maximum relative error of the verification results was 7.91 %. Through the program, the unprotected transient conditions and corresponding protection control strategy transient conditions of external load loss accident, partial loss of heat sink accident, main valve accidental closing accident and Brayton unit efficiency step decline accident are simulated. The results show that the loss of all external loads will lead to rotation speed overspeed. The introduction of negative reactivity or bypass control can effectively control the rotation speed rise, but the introduction of negative reactivity is better than bypass control. Under unprotected control, the partial loss of the heat sink, the accidental closing of the main valve and the step down of the Brayton unit efficiency will all lead to the rapid decline of the rotation speed and mass flow rate, and there is a risk of the reactor temperature exceeding the limit. The rotation speed and mass flow rate can be effectively stabilized by load following to prevent the temperature exceeding the limit.

Validation of EPG/SAG Rev. 4 of the Kuosheng Nuclear Power Plant Using the MAAP5 Code

The Boiling Water Reactor Owner’s Group released Emergency Procedure Guidelines and Severe Accident Guidelines Revision 4 (EPG/SAG Rev. 4) in 2018. The major improvement to EPG/SAG Rev. 4 was Contingency 1 (Alternate Level/ Pressure Control). Contingency 1 coordinates the reactor pressure vessel (RPV) water level and RPV pressure control action to prolong the availability of steam-driven injections and optimize the transfer to motor-driven systems. In this study, the effectiveness of the EPG/SAG Rev. 4 Contingency 1 strategy was compared with those of EPG/SAG Revision 2 Contingency 1 and Specific Major Incident Guidelines (SMI) using the Modular Accident Analysis Program, Version 5 (MAAP5). SMI was developed by the Taiwan Power Company to mitigate a Fukushima-like accident. The surrogate plant that analyzed is the Kuosheng Nuclear Power Plant (NPP). The Kuosheng NPP is BWR-6 Mark-III containment. MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic risk assessments and for verification and validation of mitigation actions specified in severe accident management guidelines. The simulation scenarios were extended loss of alternative-current power and loss of ultimate heat sink. The low-capacity, motor-driven portable pump was the only available system for RPV injection in the first hour of the accident. In this time period, the RPV water level and pressure were controlled by reactor core isolation cooling and safety relief valves. After this study, the strategy of EPG/SAG Rev. 4 Contingency 1 was successfully validated, and the effectiveness of minimum pre-depressurization RPV water level and the low-capacity, motor-driven portable pump were also demonstrated in this study.

Resilience of aerobic methanotrophs in soils; spotlight on the methane sink under agriculture.

Aerobic methanotrophs are a specialized microbial group, catalyzing the oxidation of methane. Disturbance-induced loss of methanotroph diversity/abundance, thus results in the loss of this biological methane sink. Here, we synthesized and conceptualized the resilience of the methanotrophs to sporadic, recurring, and compounded disturbances in soils. The methanotrophs showed remarkable resilience to sporadic disturbances, recovering in activity and population size. However, activity was severely compromised when disturbance persisted or reoccurred at increasing frequency, and was significantly impaired following change in land use. Next, we consolidated the impact of agricultural practices after land conversion on the soil methane sink. The effects of key interventions (tillage, organic matter input, and cover cropping) where much knowledge has been gathered were considered. Pairwise comparisons of these interventions to nontreated agricultural soils indicate that the agriculture-induced impact on the methane sink depends on the cropping system, which can be associated to the physiology of the methanotrophs. The impact of agriculture is more evident in upland soils, where the methanotrophs play a more prominent role than the methanogens in modulating overall methane flux. Although resilient to sporadic disturbances, the methanotrophs are vulnerable to compounded disturbances induced by anthropogenic activities, significantly affecting the methane sink function.

Transient analysis of the safety characteristics on a super carbon-dioxide cooled micro modular reactor

The supercritical CO2 (S-CO2) demonstrates great potential as a coolant for nuclear reactors due to its low critical pressure and temperature. It offers improved thermal efficiency and decreased requirements for structural materials. By incorporating core physics models and updated thermal–hydraulic models, the SYST-SCO2 code developed in prior research has been enhanced and utilized for conducting transient safety analysis on the micro modular reactor (MMR) proposed by the Korea Advanced Institute of Science and Technology (KAIST). The analyzed transients include the Unprotected Loss of Flow Accident (ULOFA), Unprotected Reactivity Insertion Accident (URIA), Unprotected Loss of Heat Sink (ULOHS), Unprotected Over Cooling (UOC), and the Protected Loss of Flow Accident (PLOFA). It is found that the MMR experiences the exceeding of coolant outlet temperature limit of 676℃ in the ULOFA for over 155 s. With the reactor shutdown system put into operation, the reactor core can be maintained to operate with the fuel pellet, cladding and coolant temperatures below their limits respectively in the LOFA. Enough safety margin can be left for the MMR reactor core in URIA, ULOHS and UOC transients. Among the safety limits for the fuel and coolant, the core outlet temperature limit is the easiest to reach in the analyzed accidents.

Real-time Characterization Model of Carbon Emissions Based on Land-use Status: A Case Study of Xi'an City, China

The traditional carbon accounting method, with a lag of over 2 years due to the release time of statistical yearbooks, impedes timely policy adjustments in urban planning and management. Hence, there is an urgent need to establish a real-time carbon emissions characterization model. Xi'an which has a complex land-use structure was chosen as the study site and its carbon emissions were calculated using the Emission Factor Method. The GIS-Kernel Density (KD) model was constructed, and land use was subdivided based on Point of Interest (POI) and road network data. Based on the results of carbon emissions accounting and land-use subdivision, a Multilayer perceptron (MLP) model was established. The remote sensing (RS) images of Xi'an underwent supervised classification, and the carbon emissions of Xi'an were characterized based on the subdivision results and MLP model. The results show that: (1) The accuracy of the characterization model is more than 90%, and with the improvement of RS technology, the accuracy will be further improved; (2) Compared with the existing model, this model can real time reflect the spatial distribution of carbon emissions; (3) Atmospheric emission of Xi'an will be 41.92 million tons at the end of 2022, a decrease of 2.80 million tons compared with that of 2020, but an increase of 0.33 million tons from 2021. The north of Xi'an and periphery of the central urban area are the main carbon sink loss areas, while the east of Xi'an and north foot of the Qinling Mountains are carbon sink growth areas.

A Whole-Core Transient Thermal Hydraulic Model for Fluoride Salt-Cooled Reactors

Abstract A thermal hydraulic model is developed for a solid pin-fueled fluoride-salt-cooled small modular advanced high temperature reactor (SmAHTR). This preconceptual SmAHTR was developed by the Oak Ridge National Laboratory (ORNL). For the fuel assembly configuration investigated in this study, the fuel and non-fuel pins are arranged in a hexagonal layout. The molten FLiBe salt coolant flows parallel to the bank of pins. A finite volume model is developed and used to compute temperatures in the solid regions (fuel and non-fuel pins, and the graphite reflectors) in the core. The temperature, flow, and pressure profiles for the coolant flowing through the pin bundles in the core are calculated using the conventional subchannel methodology. Pertinent closure relations are used to compute the hydraulic losses, momentum, and energy exchange between adjacent subchannels, and heat transfer between the solid and fluid regions. The resulting model can perform both steady-state and transient computations across the entire core. This fully implicit model also includes an adaptive time-stepping algorithm for automatic time-step adjustment. A preliminary code-to-code comparison demonstrates good agreement between the present subchannel-based model and a computational fluid dynamics (CFD)-based model for a transient case in which the core inlet flowrate varies with time. Following the code-to-code comparison, the thermal hydraulic model is used to analyze the protected loss of heat sink (P-LOHS) accident scenario. Key results such as the transient evolution of peak fuel, non-fuel, and coolant temperatures, and 3-D core temperature distribution are presented and discussed.

Risk reduction by means of FLEX strategies in pressurized water reactors

One of the key lessons learned from the Fukushima Dai-ichi accident was the importance of the challenge presented by the loss of safety-related systems after a Beyond Design-Basis External Event (BDBEE). In particular, Extended Loss of AC Power (ELAP) or Loss of Ultimate Heat Sink (LUHS) can severely compromise key safety functions associated with core cooling and containment, ultimately leading to reactor core damage. Because of that the so-called Diverse and Flexible Coping Strategies (FLEX) were designed.These aforementioned FLEX strategies are also useful to avoid core damage in sequences due to internal events. Therefore, these FLEX strategies should be incorporated in the Level 1 Probabilistic Safety Assessment (L1PSA) to analyse their impact on core damage frequency (CDF). This work quantifies the impact of FLEX strategies by using a generic model (3 loops, PWR-WEC) developed by the Universidad Politécnica de Madrid (UPM) in collaboration with the Spanish Nuclear Regulatory Body (CSN). As a result, a decrease of the CDF at power for internal events is observed and, therefore, indicates that these FLEX strategies have a positive impact on the plant safety not only for BDBEE but also for other accidental conditions.

Scenario analysis loss of heat removal from the spent fuel pool on nuclear power plant

The spent nuclear fuel storage system is designed to store and cool spent fuel for several years taking into account scheduled reloads and unloading of the entire core accumulated after use in a nuclear reactor. It consists of special pools or containers where spent fuel is placed for temporary storage before final treatment or disposal. These systems provide safe and efficient storage of spent fuel to prevent radioactive material from leaking into the environment and minimize risks to human health and the natural environment. The events that occurred during the Fukushima nuclear disaster on March 11, 2011, underscored the importance of safe storage of spent fuel in the spent fuel storage pool. Storage safety has therefore become a key aspect in this area. This article describes the heat sink loss calculations for the analytical substantiation of the emergency response instructions for the shutdown state of the Armenian NPP power unit No.2 using the RELAP5/Mod3.2 computer code. The initiating event in the event of loss of heat removal from the spent fuel pool is considered. The analysis of nuclear safety in the course of the development of a beyond design basis accident with a long NPP blackout was carried out in relation to the spent fuel pool of a power unit with a reactor plant WWER-440 (project V-270). The radiation consequences are estimated. The article provides calculations of the following accidents to determine the necessary actions of the operator: loss of heat removal from the spent fuel pool without operator action and loss of heat removal from the spent fuel pool from the organizations of subsequent make-up of the spent fuel pool with a boron cleaning pump (2NBO-2). The calculations are based on boundary and initial conditions corresponding to the assumptions of the «better estimate».

Improved photosynthetic performance under unilateral weak light conditions in a wide-narrow-row intercropping system is associated with altered sugar transport.

Intercropping improves resource utilization. Under wide-narrow-row maize (Zea mays) intercropping, maize plants are subjected to weak unilateral illumination and exhibit high photosynthetic performance. However, the mechanism regulating photosynthesis under unilateral weak light remains unknown. We investigated the relationship between photosynthesis and sugar metabolism in maize under unilateral weak light. Our results showed that the net photosynthetic rate (Pn) of unshaded leaves increased as the level of shade on the other side increased. On the contrary, the concentration of sucrose and starch and the number of starch granules in the unshaded leaves decreased with increased shading due to the transfer of abundant C into the grains. However, sink loss with ear removal reduced the Pn of unshaded leaves. Intense unilateral shade (40% to 20% normal light), but not mild unilateral shade (60% normal light), reduced grain yield (37.6% to 54.4%, respectively). We further found that in unshaded leaves, Agpsl, Bmy, and Mexl-like expression significantly influenced sucrose and starch metabolism, while Sweet13a and Sut1 expression was crucial for sugar export. In shaded leaves, expression of Sps1, Agpsl, and Sweet13c was crucial for sugar metabolism and export. This study confirmed that unshaded leaves transported photosynthates to the ear, leading to a decrease in sugar concentration. The improvement of photosynthetic performance was associated with altered sugar transport. We propose a narrow-row spacing of 40 cm, which provides appropriate unilateral shade and limits yield reduction.

Heat Transfer Performance of Water-cooled Heat Sink for I&C System Controller Chip in Nuclear Power Plant

To improve the heat dissipation reliability of the nuclear power control system chip, the structure of its water-cooled heat sink is compared and analyzed. Three schemes of water-cooled heat sinks are proposed, and CFD analyzes the heat dissipation performance of the heat sink. The simulation data of the heat transfer coefficient, pressure loss, and heat sink comprehensive coefficient are compared. The results show that the average temperature of the bottom surface of the three heat sinks tends to be flat with the flow velocity increase, and the heat dissipation coefficient Nu is also the same. The S-type heat sink has a higher heat transfer coefficient and lower average bottom temperature. The pressure loss of the S-type heat sink is much higher than that of the other two schemes with the increased flow rate, and the system resistance is large. By calculating the comprehensive coefficient of the heat dissipation process of the three heat sinks and considering the heat dissipation capacity and other factors, it is considered that the circular fin heat sink can be the optimal solution.

Protected Loss-of-Flow Accident in a HLM Pool Facility: Long-Term Cooling Experiment

Since the 2000s, the development of Generation-IV fast reactors cooled by heavy liquid metals (HLMs) has been pursued by several research activities and projects, many of which are co-funded by the European Commission. One of the key points of HLMs regards their good neutronic and thermophysical properties, allowing for the of design cores with a high pitch-to-diameter ratio. In terms of passive safety, with a properly designed configuration, it is possible to increase the system capability to remove the decay power in a natural circulation regime, reducing active safety systems involvement. Such a safety-related aspect has been experimentally investigated at the ENEA Brasimone Research Center within the European Union co-funded Euratom H2020 SESAME project. An experimental campaign reproducing protected loss-of-flow accident (PLOFA) scenarios has been executed on CIRCE, a lead-bismuth eutectic-cooled pool-type facility reproducing in relevant scale the main components of HLM-cooled fast reactors. A test section named HERO has been installed in the CIRCE main vessel, hosting a double-wall bayonet tube steam generator scaled 1:1 in length with respect to the one envisioned for the ALFRED reactor. The tests reproduce the loss of primary coolant flow, while the power supplied by the heating source decreases according to a characteristic decay heat curve. The feedwater in the secondary system is regulated to operate the main steam generator as a decay heat removal (DHR) system. This paper presents the PLOFA transient reproducing the worst case, where the steam generator feedwater is suddenly stopped, simulating the full loss of the heat sink (no DHR). The main phenomena occurring during the transition from forced to natural circulation are presented and discussed. The experiment shows that, despite the loss of the forced circulation regime in the primary loop and the full loss of the heat sink, the entire system is still capable of operating safely, assuring an effective long-term cooling, as long as the thermal heat losses from the main vessel balance the decay power supplied by the heating source.

Research on safety characteristics of heat pipe reactor with semiconductor thermoelectric conversion system

The heat pipe reactor exhibits notable characteristics of inherent safety and a compact configuration, rendering it a favored choice for space and ocean reactor designs. This study focuses on investigating the safety characteristics of the SAIRS-C space heat pipe reactor by employing a self-developed TAPIRS-D code under three selected typical accidents. The obtained results demonstrate that the heat pipe reactor exhibits inherent safety features, enabling it to sustain normal operation even in the event of partial heat pipe failure(2 out of 60 heat pipes fail) and partial heat sink loss(cooling rate decreasing from 100% to 10% in 10 s). The reactivity-insertion accident(0.358$ is introduced into the core) is found to exert the most significant influence on the heat pipe reactor. However, the crucial parameters of the reactor during the accident, such as fuel temperature, remain within the established safety limits. The outcomes of this research not only contribute to a comprehensive understanding of safety characteristics of heat pipe reactor, but also provide a crucial guidance for further design enhancements.

Transient analyses for China initiative Accelerator Driven System using the extended BELLA code

China initiative Accelerator Driven System (CiADS) is a 10 MW lead–bismuth eutectic (LBE) cooled subcritical reactor loaded with UO2 fuels in the first phase, which is designed to demonstrate the engineering feasibility of the ADS concept. The transient analyses were performed to show the safety potential and dynamic characteristics of the CiADS subcritical reactor using the extended BELLA code. Typical scenarios, such as beam-trip transients, unprotected beam overpower (UBOP), unprotected and protected loss of flow (ULOF and PLOF), unprotected and protected loss of heat sink (ULOHS and PLOHS) and self-defined station blackout (SBO), were analyzed and simulated. The results indicate that, as a low-power reactor, the CiADS subcritical reactor has a large margin to avoid severe damage in the core. Short-term cladding failure caused by thermal fatigue under frequent beam trips may not happen since the variations of temperatures are relatively small. However, accelerated LBE corrosion combined with accumulated creep tends to be a risk under ULOF and ULOHS.

Thermal-Hydraulic Transient Analyses for Primary Coolant System of HFRR Using RELAP5

The High Flux Research Reactor (HFRR) is a 40 MW pool-type nuclear research reactor developed for neutron beam tube research, material and fuel irradiation studies, and advanced research using a cold neutron source. As a part of safety analyses, thermal-hydraulic transient analyses were carried out using the computer coder RELAP5 for the primary coolant system of the HFRR. This technical note presents various analyses that were carried out such as loss of normal electric power supply, loss of heat sink, main coolant pump seizure, inadvertent opening of a natural circulation valve, and loss-of-coolant accident considering breaks in the pump suction line and pump discharge line. It is demonstrated that fuel cooling is ensured and safety parameters are well within acceptable limits.

Impact of removal of rubber plantations for urbanization on CO2 mitigating capacity by the loss of carbon sink in Kerala state, India

Mitigating climate change and global warming through carbon sequestration by tree ecosystems is of prime importance since they are cost-effective, environmentally friendly and ecologically sustainable. Urbanization is a part of development, and rubber plantations are usually removed for this purpose, especially in Kerala, the southern state of India. Besides latex, the economic produce, and the associated income, the rubber tree is a fairly good sink for carbon in its biomass, with an average carbon content of 42 per cent and substantial carbon stock in the soil. In the present study, an account of total carbon loss by the removal of rubber plantation for urbanization and developmental activities are given. The present popular clone (RRII 105) existing in major share (85%) of the total rubber cultivation in India accounts for carbon sink loss 57 t ha-1, 57.5 t ha-1, 43.2 t ha-1 for 23 years and 148 t ha-1, 75 t ha-1 and 62.1 t ha-1 for 30 years from biomass, litterfall and sheet rubber respectively. The recent clones RRII 414, RRII 429 and RRII 417 have higher growth rates and higher biomass (44-50 per cent) carbon sink loss compared to the existing popular clone RRII 105. The carbon sink loss in the form of stored carbon in soil is 56.5, with a soil carbon content between 1.2 to 2 per cent. Due to the growth variation in diverse environments with extreme climatic conditions, the clones recorded differences in carbon stock and carbon sink loss. The central region of Kerala showed a higher loss, and a lower loss was in the drought-affected northern region than the southern region. The total carbon sink losses for 23 and 30 years were 214.2 and 341.5 t ha-1, respectively. This study points out that the serious carbon sink loss due to the removal of rubber plantations results in disturbing the self-sustained, carbon-friendly and economically sound perennial rubber ecosystem. Vegetation having higher C-sequestration potential and trees with higher lignin content is essential to increase carbon capture for mitigating the impact of the removal of plantations. From the present study, it is clear that the removal of rubber plantations is affecting the carbon sink loss, thereby the CO2 mitigating capacity, and is a serious matter of concern.

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Abstract Feedback reactivity caused by core deformation is one of the inherent safety features in a sodium-cooled fast reactor (SFR). To validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code, named Super-COPD, we have been conducting the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of balance of plant (BOP)-302R and BOP-301 as one of the representative issues in Experimental Breeder Reactor-II (EBR-II), a pool-type experimental SFR in the United States. During the transient of the ULOHS tests, the reactor power decreased to the decay heat level due to the negative reactivity caused by the radial expansion of the core. Developing the evaluation models of the reactivity feedback is essential for predicting the plant behavior in unprotected events, such as ULOHS events, and the numerical analyses modeling the feedback reactivity were conducted. By comparing the numerical results and the experimental data, the increment temperature profiles of the core inlet and the decreasing reactor power profile calculated by Super-COPD were comparable with experimental data. The applicability of the evaluation models for the reactivity feedback was indicated during the ULOHS event. In addition, through the sensitivity analyses on the cold pool model, it was found that the modification of the plenum model of the cold pool to take into account the thermal stratification was required to improve the core inlet temperature profile.

Carbon sinks/sources' spatiotemporal evolution in China and its response to built-up land expansion

Terrestrial ecosystem carbon sink examination in China still faces great uncertainties. Determinant analysis has focused on climate change but ignored the influence of fast urban expansion. Using remote sensing images, climate variable data, and high-resolution land use data, this research improved net ecosystem productivity (NEP) simulation model based on a large number of field observations, and investigated spatial-temporal changes of NEP. This research calculated the NEP loss caused by built-up land expansion and used geographically weighted regression (GWR) model to explore the spatial heterogeneity of the relationship between NEP growth and built-up land expansion. The results showed that China contributed a carbon sink of 0.33 Pg C per year from 2000 to 2020. Southern China had a greater capacity to sequester carbon than northern China. The carbon sink capacity of most Chinese regions increased. Built-up land expansion caused 4.95 Tg C of carbon sink loss per year, which was mainly concentrated in eastern China. In GWR model, 50.8% of regions showed negative correlations between NEP growth and built-up land expansion. These two variables were mostly positively correlated in the northwest and negatively correlated in the southeast. Consequently, this study suggests that maintaining the capacity of carbon sinks in southern provinces is important for China to meet its carbon neutrality goal.

Effects of fuel conversion ratio on cooling and drag reduction performance for supersonic film using gaseous hydrocarbon fuel

Film cooling especially combined with regenerative cooling is an effective solution to meet the heat protection and the drag reduction demands for hydrocarbon-fueled scramjet engines. Typically, fuel is the only available coolant to organize the active cooling for hypersonic vehicles, and the coolant used in the film cooling is extracted from the regenerative cooling channels; however, the fuel pre-cracking in the regenerative cooling process must be considered as it can change the initial composition and energy level of the coolant in the film cooling. Numerical simulations based on the RANS method are conducted to evaluate the effects of film pre-cracking on its cooling and drag reduction performance, by changing the fuel conversion ratio from 0 to 40 %. Results show that the influence of film inlet cracking on its drag reduction performance is limited, and the relative change of the wall friction under different conversion ratios is less than 5 %. However, the film inlet cracking has a notable negative influence on its thermal protection performance. Specifically, for every 10 % increase in the film conversion ratio, the average adiabatic wall temperature will rise by about 20 K. Further analysis reveals that although the total amount of the fuel chemical heat sink is comparable to its physical heat sink, only around 25 % of the film chemical heat sink can be effectively used to enhance the fuel film cooling performance. The loss of the effective chemical heat sink during the pre-cracking process is the main reason for the decrease of film cooling performance as the film inlet conversion ratio increases.