Coolant Circuit Research Articles

Self-oscillations that cause the destruction of intermolecular bonds in the water coolant of a nuclear power plant

The digital acoustic model of a nuclear reactor (DAMNR) is presented as an auto-oscillatory system belonging to a special class of nonlinear dissipative systems capable of generating undamped oscillations. It is established that a water-water power reactor with a turbulent flow of a coolant is an open system of high complexity with a large number of elements, the connections between which are not predetermined, but probabilistic. Elements of the coolant circuit with negative dissipation (negative friction) are identified. It is shown that they self-organize chaotic turbulent pulsations and vortices into ordered wave oscillations, the frequency of which is determined by the Thomson (Kelvin) formula. In radio engineering circuits, an electronic self-oscillating generator with transformer feedback has similar properties. The presence of negative resistance in nonlinear dynamical systems leads to self-organization of chaotic turbulent perturbations and generation of self-oscillations in the form of acoustic standing waves (ASW). On the basis of theoretical and experimental data, the reliability of a previously unknown property of a reactor with connected pipelines - the ability to generate several ASW simultaneously-was confirmed. The use of DAMNR in the design and operation of nuclear power plants allows to identify the sources of ASW occurring in the coolant, the conditions for their occurrence and frequency.

Simulation of mixing flows with different temperatures in a T-junction using OpenFOAM code

This paper aims to evaluate the frequency of velocity and temperature fluctuations in the mixing region using OpenFOAM code. Turbulent mixing of fluids at different temperatures can lead to temperature fluctuations at the pipe material. These fluctuations, or thermal striping, inducing cyclical thermal stresses and resulting thermal fatigue, may cause unexpected failure of pipe material. Therefore, an accurate characterization of temperature fluctuations is important in order to estimate the lifetime of pipe material. Thermal fatigue of the coolant circuits of nuclear power plants is one of the major issues in nuclear safety. To investigate thermal fatigue damage, the OECD/NEA-Vattenfall T-Junction Benchmark was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to predict the important parameters affecting high-cycle thermal fatigue in mixing tees. In this study, to simulate the standard problem described above, the OpenFOAM code is used, which is an open integrated platform for numerical simulation of problems in continuum mechanics. At the first with Salome-meca code, a computational grid was created, consisting of about 450,000 nodes, and k-eps model and RANS models were used to simulate turbulence. OpenFOAM code results were compared with the available experimental results. The results were found to be in well-agreement with the experimental results in terms of amplitude and frequency of temperature and velocity fluctuations.

Corrosion Phenomena of High Temperature Alloys by Helium Impurities in the Primary Coolant of HTGR

The primary coolant circuit of the high temperature gas-cooled reactor (HTGR) contains trace impurities. A nickel base alloy would corrode when exposed to an atmosphere at a high temperature and for a long time. The protective oxide scale formed by chromium is an important factor to prevent severe corrosion of high temperature alloys. Corrosion tests were conducted on Inconel 617, Incoloy 800H, Hastelloy X, and T-22, which are commonly used in the steam generator of HTGR. The alloys were exposed to helium with trace impurities for 48 hours at 950°C. The corrosion results were analyzed by weighing, scanning electron microscopy (SEM) and electron probe microanalyzer (EPMA). All the four alloys formed oxide scales in this atmosphere, but they differ in the capacity to resist corrosion. Therefore, the carbon transfer phenomenon observed in this experiment varies for the different alloys. In addition, for Cr in Inconel617, the expected depletion phenomenon near the corrosion layer occurred, which is consistent with the results from theoretical analysis.

The influence of MHD boundary layers on tritium permeation in PbLi flows for fusion breeding blankets

In PbLi based breeding blanket concepts, tritium is produced inside the liquid metal and drag out of the reactor by the liquid metal flow. However, undesired permeation through the channels and pipes walls occurs spontaneously since tritium naturally diffuses in the opposite direction of the concentration gradient. This way tritium can reach the blanket coolant circuit or even the exterior with an impact on the tritium self-sustainability and the safety of the plant. Similarly to heat transfer processes, permeation through the walls in the interface between the flow and the steel is mostly affected by the dynamics of the boundary layers. This is ruled by the electrical coupling between the moving conductor and the conducting walls as a result of the Magnetohydrodynamics (MHD) interactions which dominate the flow dynamics. In this work, the connection between the MHD forces and tritium transport is numerically studied using the simulation platform ANSYS-Fluent. The velocity profiles of a PbLi test channel have been firstly computed in a wide range of Hartmann numbers from 102 to 104. These velocity profiles are then applied to a 3D tritium transport model developed with the customization capabilities of the same platform. A series of tritium transport simulations are carried out considering different permeation regimes: surface-limited, diffusion-limited and intermediate regimes. The development of the concentration boundary layers along the channel is studied in different permeation regimes, magnetic fields and velocity fields. This has allowed correlating the Sherwood number (Sh) with the Hartmann (Ha), Reynolds (Re) and permeation numbers (W).

Metamodeling and On-Line Clustering for Loss-of-Flow Accident Precursors Identification in a Superconducting Magnet Cryogenic Cooling Circuit

In the International Thermonuclear Experimental Reactor, plasma is magnetically confined with Superconductive Magnets (SMs) that must be maintained at the cryogenic temperature of 4.5 K by one or more Superconducting Magnet Cryogenic Cooling Circuits (SMCCC). To guarantee cooling, Loss-of-Flow Accidents (LOFAs) in the SMCCC are to be avoided. In this work, we develop a three-step methodology for the prompt detection of LOFA precursors (i.e., those combinations of component failures causing a LOFA). First, we randomly generate accident scenarios by Monte Carlo sampling of the failures of typical SMCCC components and simulate the corresponding transient system response by a deterministic thermal-hydraulic code. In this phase, we also employ quick-running Proper Orthogonal Decomposition (POD)-based Kriging metamodels, adaptively trained to reproduce the output of the long-running code, to decrease the computational time. Second, we group the generated scenarios by a Spectral Clustering (SC) employing the Fuzzy C-Means (FCM), in order to identify the main patterns of system evolution towards abnormal states (e.g., a LOFA). Third, we develop an On-line Supervised Spectral Clustering (OSSC) technique to associate time-varying parameters measured during plant functioning to one of the prototypical groups obtained, which may highlight the related LOFA precursors (in terms of SMCCC components failures). We apply the proposed technique to the simplified model of a cryogenic cooling circuit of a single module of the ITER Central Solenoid Magnet (CSM). The framework developed promptly detects 95% of LOFA events and around 80% of the related precursors.

Разработка алгоритма определения неисправности измерительных каналов системы акустического контроля течей

Nowadays Russian and foreign NPP operate systems for monitoring coolant leakage at the primary circuit based on measuring the dispersion of generated acoustic signals (acoustic waves) propagating over the metal surface. In the acoustic leak monitoring systems provision are made to self-diagnosis of measuring channels, as well as an adaptive algorithm is applied that allows automatic readjusting to the use of neighboring measuring channels instead of those that have failed. At the same time, there may be such malfunctions in the system technical means that do not automatically diagnose the malfunction of the measuring channels, which may lead to the failure of the system function to determine the magnitude and coordinate of the leak of the primary circuit coolant. That is why, the task of developing algorithms for determining the malfunction of the measuring channels of the acoustic leak monitoring system, implemented using software without making changes to the technical means of the system, is urgent. An algorithm is proposed for determining the malfunction of the measuring channels of the acoustic leak control system using a test signal of increased duration. An analysis of the applicability of the algorithm was performed on a representative sample of signals from measuring channels of the acoustic leak monitoring system of campaign 2018--2019 Novovoronezh NPP-2 рower Unit no. 1. The proposed algorithm has been implemented in the testing mode at power Unit no. 1 of Novovoronezh NPP-2 since the start of a new campaign in July 2019

Evolutionary Design of Novel Coolant Passages for Cooling a Square Substrate by Single Stream

Abstract Different designs of novel coolant (i.e., water) circuits have been proposed using a well-established constructal law to cool a square substrate made up of aluminum oxide and subjected to a uniform wall heat flux (i.e., q″=200 W/m2) at its top. Five different flow-path topologies: case-1 (umbrella-shaped), case-2 (dumbbell-shaped), case-3 (hexagonal-shaped), case-4 (down-arrow-shaped), and case-5 (up-arrow-shaped) are evolved from a single pipe embedded in the heated substrate. The best cooling pathway has been anticipated by comparing the thermo-fluid characteristics of all the designs. A numerical route, via ansys R 16, has been implemented to solve the transport equations for continuity, momentum, and energy along with relevant boundary conditions. To access a better design, the nondimensional temperature and pressure drop for these cases have been quantified and compared, by varying the length and Reynolds number in the range of 2≤Lc/L≤3 and 100≤Re≤2000, respectively. We observe a decrease in the temperature and an increase in the pressure drop with Reynolds number for all the considered pathways. When Re≤500, a rapid fall in the nondimensional temperature has been noticed; and thereafter, it looks like a plateau for all cases. For case-4, a minimum temperature is obtained at the nondimensional pipe length of 2.5. At Lc/L≤2.5, we observe that the case-4 provides better cooling to the substrate among all other designs. Also, the pressure drop for case-4 is not too high as compared to other designs.

Pressurizer water level control with estimation of primary circuit coolant mass in nuclear power plants via robust observer based dynamic sliding mode control

Pressurizer is one of the most important components in a pressurized water nuclear power plant. Since it is responsible for coolant mass balance and non-boiling heat transfer in the primary circuit. The pressurizer water level control system in a nuclear power plant with the Pressurized Water Reactor (PWR) is responsible for the coolant mass balance. The main control goal is to stabilize the water level at a reference value and to suppress the effect of time-varying disturbances such as the coolant leakage in the primary circuit pipeline system. In the process of PWR power plant operation, the incorrect pressurizer water level may disturb pressure control or may cause damage to electric heaters which could threaten plant security and stability. In modern reactors, standard Proportional–Integral–Derivative (PID) controllers are used to control the water level in a pressurizer. In this paper, for the first time, in order to improve the performance and transient response of the pressurizer control system, an observer-based dynamic sliding mode control using Lyapunov-approach is designed and applied for the level control problem of the pressurizer in a pressurized water nuclear power plant based on the validated nonlinear model. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. The comparison between the Dynamic Sliding Mode Control (DSMC) and the PID controller shows a significant improvement in water set-point tracking and increased ability in disturbance rejection for the proposed observer-based DSMC. Simulation results confirm the improvement in the transient response obtained by using the proposed controller.

Thermo-hydraulic analysis of PbLi ancillary system of WCLL TBM undergoing in-box LOCA

The European liquid Test Blanket Module (TBM) concept named Water Cooled Lithium Lead makes use of the liquid metal alloy PbLi as tritium breeder/neutron multiplier and water as coolant. A dedicated PbLi ancillary system feeds the detritiated liquid metal alloy to the TBM and returns the PbLi enriched with tritium. The object of this study is developing a numerical model able to simulate the thermo-hydraulic behavior of the PbLi ancillary system under both normal operation and accidental scenario, with the aim to investigate the propagation of pressure waves generated by an in-box LOCA. The accident consists of a tube rupture of the coolant circuit inside the TBM, leading to the interaction between PbLi and water. As a consequence, a pressure increase occurs inside the breeder zone. The accident might affect the ancillary systems due to the propagation of the pressure waves generated inside the TBM. The PbLi loop is modelled by means of the thermo-hydraulics system code RELAP5/Mod3.3 and an instantaneous pressure peak is applied inside the control volume simulating the TBM soon after reaching the steady state conditions inside the whole loop. The propagation of the pressure wave inside the circuit and the hydraulic behavior of the individual components is studied. The present paper represents a first step towards the investigation of the PbLi system and its safety features with the aim to support the R&D activities carried out in the framework of the ITER project.

Carbon Dioxide Contamination of Aqueous Morpholine Solutions and Effects on Secondary Coolant Chemistry Under CANDU Conditions

Exposure to air can cause amine solutions in CANada Deuterium Uranium (CANDU) reactor secondary coolant circuit feed tanks to absorb carbon dioxide (CO2). Likewise, carbon dioxide can be absorbed directly into the amine-containing secondary coolant by air ingress during shutdown, lay-up, and startup. Sampling operations, including transferring the sample to the laboratory and subsequent analyses, can also provide opportunities for CO2 contamination. This paper reports the results of laboratory and chemical modeling studies to examine the effects of CO2 contamination on aqueous morpholine solutions. The chemistry of CO2 uptake by feed tanks containing up to 50 wt% (11.5 mol·kg−1) morpholine at 25°C was modeled using the OLI Studio 9.5.2 chemical equilibrium model, and the speciation was confirmed by 13C nuclear magnetic resonance spectroscopy (NMR) measurements. The effects of CO2 contamination on the pH of the secondary coolant containing 60 ppm (0.006 wt%, 7.00 × 10−4 mol·kg−1) morpholine and the resulting effects on the solubility of magnetite and nickel oxide from 25°C and 250°C at steam saturation were modeled as a function of CO2 loading using the Electrical Power Research Institute chemical modeling software MULTEQ v.8. The chemical modeling calculations show that concentrated alkaline morpholine solutions at room temperature and pressure would be expected to have a strong tendency to absorb CO2 and have additional uptake abilities due to the formation of morpholine carbamates. For dilute morpholine solutions at room temperature and pressure, the solutions are still sufficiently alkaline to absorb enough CO2 to cause a measurable change in the pH of the secondary coolant. This effect was shown to be negligible under reactor operating conditions. The absorption of CO2 would potentially have the most effect on either unprotected feed tanks or during lay-up conditions in the steam generators, as it could depress the pH of the lay-up solution and adversely affect the rate of corrosion in the internal components of the steam generators (e.g., carbon steel materials). The 13C NMR measurements on samples of 50 wt% aqueous morpholine solutions from feed tanks at the Ontario Power Generation’s Pickering Nuclear Generating Station found that CO2 was below the 0.02 wt% detection limit, and suggest that the procedures used to avoid CO2 contamination in feed tanks are effective. The 13C NMR was shown to be an effective tool for monitoring CO2 uptake by morpholine solution in the feed tanks under conditions in which they may have undergone abnormal exposure to air.

Preliminary analysis of an in-box LOCA in the breeding unit of the WCLL TBM for the ITER reactor with SIMMER-IV code

The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (Water-Cooled Lead-Lithium Breeding Blanket) that will be tested in the Test Blanket Module of the ITER reactor. In this accident, a tube rupture in the coolant circuit would cause a potential disruptive exothermic chemical reaction between the water (the primary coolant) and the Lead-Lithium. This paper presents the development of numerical tools for the analysis of the in-box LOCA consequences in the breeding unit of the blanket.In this work a simplified computational model of the breeding unit was created by SIMMER-IV which includes the lead lithium and water systems and the internal solid structures. The simulation is set up at the foreseen nominal working conditions of the units and involves a small rupture of the cooling tube, with the consequent propagation of high pressure waves on the breeder side and the fast evolution of the chemical reaction. It is shown here that the code is capable of correctly predicting both these events, from a qualitative point of view, thus providing a highly valuable insight of the whole scenario. These preliminary results will provide a crucial base for further development of this computational tool and the simulation of the whole Test Blanket Module.

Integration of DEMO hazard piping into the tokamak building

Because of the complexity of the design of a Fusion Power Plant like the EU DEMOnstration power plant the study of the plant layout must proceed in parallel with the design of the major systems. This is necessary to ensure that the design of the plant incorporates from the very beginning sound considerations on:•Safety and licensing,•Personnel security in context with operation and maintenance,•Adequate plant availability, and•Safe performance operation including delivery of few hundred MW net electric power to the electrical grid.Though it is a process of trial and change that follows the design evolution, this approach allows a better and continuous control of the numerous physical and functional interfaces among the systems and structures assuring an optimization of the overall design focused on the above targets.Some of the fluid systems inside the tokamak building are activated e.g., water coolant circuits of in-vessel components containing activated corrosion products (ACPs), radioactive N16 and N17 isotopes generated from neutron irradiation of oxygen, as well as the Lithium-Lead (LiPb) system containing ACPs and radioactive isotopes generated by neutron activation of the LiPb alloy. The layout of the corresponding circuits must consider constraints such as shielding, inspection, maintainability and irradiation lifetime of some equipment of other plant systems located in the tokamak building, e.g. electric and electronic equipment, organic seals of valves, and relevant actuators. Furthermore the personnel exposure during operation and maintenance has to be as low as reasonable achievable.The experiences of Nuclear Power Plants and ITER (International Thermonuclear Experimental Reactor) are an important input to the layout of the DEMO coolant circuit. This paper presents some initial considerations on the criteria to be used for the layout criteria inside the tokamak building.

Innovation Control Strategy Simulation Research of Once-through Steam Generator

The new type of nuclear power plant adopts once-through steam generator (OTSG) which can supply steam to the turbine, control and operation of OTSG is different from the traditional natural circulation steam generator, the control strategy of the traditional natural circulation steam generator is not applicable to OTSG. In order to meet the demand of the general operation of the integrated reactor, the operation control of OTSG is put forward. The paper proposes a double const control strategy that the average temperature of the primary circuit coolant and the secondary circuit steam pressure are kept constant, also proposesa coordinated control strategy considering reactor and turbine as a whole. The control strategy is different the traditional control method. In order to verify the control strategy, simulation experiments are carried out in a full-scale simulator system. The simulation results show that the proposed control strategy of OTSG in integrated reactor to meet the load demand, also can effectively guarantee the dynamic characteristics of pressure steam generator and related parameters of OTSG.

Numerical investigation of the tritium permeation phenomenon through cooling plates in breeding blankets

Understanding and predicting tritium transport is considered a key issue for the design of safe and self-sufficient fusion power plants. Indeed, tritium generated in the breeder materials is susceptible to permeate to the blanket coolant circuits where it is harder to remove and process. Therefore, the predictive models become very important to evaluate the performance of any blanket design. In this work, 2D finite volume methods are employed to analyze in detail the tritium permeation through a cooling plate considering a set of different parameters that affect the breeding blanket (BB) performance. Different BB configurations are studied varying the geometrical parameters defining the plate: the wall thickness and the cooling channel pitch. Moreover, the analyses cover every permeation regime, from diffusion-limited to surface-limited, including intermediate situations. This way, results are applicable to every BB concept which employs cooling plates with independence of their specific materials, temperatures and tritium concentrations. Results have been compared with those of 1D models in order to define form factors that can be used to increase the precision of system level tritium transport models. A total of 91 simulations have been carried out to find correlations that allows computing the form factors in a wide variety of situations. Results show that the accuracy of 1D models and consequently the accuracy of most system level models strongly depends on the permeation regime and can be poor in some conditions. However, this study demonstrates that the use of form factors can be used for an effective increase of system level exactness.

Thin-film distillation coupled with membrane condenser for brine solutions concentration

A novel concept of film distillation equipped with the membrane condenser (FD-MC) was proposed and implemented for the concentration of brine salt solutions. In this thermo-gradient method, the water was evaporated from the thin liquid film flowing alongside the hot surface, and then directly condensed on the cold surface of the porous membrane placed at the distance of few millimeters from the feed. The change of the membrane positioning from the hot feed solution to the cold water stream (coolant circuit) enables to overcome the common challenges of the membrane distillation process like long-term stability towards pores wetting, membrane scaling, and latent heat loss. Comparing with the air gap membrane distillation with membrane condenser (AGMD-MC), FD-MC demonstrated higher water flux (14.7 kg/m2·h) and stable performance during the concentration of NaCl solution from 50 up to 230 g/kg (microcrystal formation was noticed) even in the presence of organic pollutants (kerosene or surfactants). A mathematical model of heat and mass transfer in FD-MC process was proposed and successfully verified. With the respect to the temperature of hot (40–100 °C) and cold (10–60 °C) circuits, the energy consumption and thermal efficiency of FD-MC process were in the range of 2.7–3.2 MJ/kg and up to 97%, respectively.

New designs of target and cooling scheme for water cooled divertor in DEMO

Developing a robust and reliable divertor with an adequate heat removal capability is one of the most crucial issues for the realization of DEMO. Although lots of designs and technologies have been accumulated for the water cooled divertor and a full tungsten water cooled divertor for ITER has been successfully developed, there is still a gap to catch up the requirements of DEMO mainly because of the neutron irradiation. For the purpose of putting forward a feasible proposal of the divertor for DEMO concept in a short to medium term, new designs for the target and cooling circuit have been explored in this article. According to the distribution characteristics of heat load and neutron irradiation damage on the divertor and the unique properties of materials, CuCrZr and RAFM are chosen as the heat sink material for the plasma facing unit (PFU) at strike zone and other zone respectively, which is named ‘dual PFUs’. A single coolant circuit is applied to both PFUs and a new design ideology of the cooling scheme has been developed. To assess the performance of the designs, corresponding thermal and structural response for the PFUs and thermal-hydraulics analysis for the cooling circuit had been performed. The property of CuCrZr in neutron irradiation environment is the main limitation for the performance of the target. And the new design for cooling circuit can effectively reduce the pressure drop without any negative impact on the divertor by rationally allocating the coolant through outer vertical target to inner vertical target and dome.

Identification of LOFA precursors in ITER superconducting magnet cryogenic cooling circuit

In the International Thermonuclear Experimental Reactor, plasma is magnetically confined with Superconductive Magnets (SMs) that must be maintained at cryogenic temperature by a Superconducting Magnet Cryogenic Cooling Circuit (SMCCC). To guarantee cooling, Loss-Of-Flow Accidents (LOFAs) in the SMCCC are to be avoided. In this work, an approach to identify LOFA precursors (i.e., those component failures leading to a LOFA) is presented. The approach is based on a Spectral Clustering (SC) method using the Fuzzy C-Means (FCM) algorithm and is applied to the SMCCC of a single module of the ITER Central Solenoid (CS).

Scientific bases of creation and practical application of digital acoustic model of NPP with VVER

Created acoustic model of nuclear reactor is presented as self-oscillating system belonging to special class of nonlinear dissipative systems capable of generating undamped oscillations with parameters that don’t depend on the initial conditions and are determined only by the properties of the system itself. In open systems, such as coolant circuit, the energy increases over time due to negative friction (negative dissipation). It is known that self-organization, in which a wave order arises spontaneously from turbulent disorder, is possible due to the exchange of energy and mass with the environment, when dissipative losses are compensated from the outside. The creation of this model combined three fundamental scientific results: the Thomson-Kelvin formula for discharge of capacitor in 1853; the properties of acoustic resonator, formulated by Helmholtz in 1869; proof by the author in 1984 of validity of the use of the method of electroacoustic analogies for study of pulsating flows of single-phase and two-phase medium in the presence of negative friction. The acoustic model of nuclear reactor has no analogues in acoustics and is designed to predict, diagnose and prevent the occurrence of unwanted self-oscillations of coolant and resonances with fluctuations of nuclear fuel and equipment in operational and emergency modes.

An improved quality assessment of fuel pin end plug welds using digital X-ray radiography

In fast breeder reactors, the quality of fuel pin end plug welds is very important because the presence of any weld defect can cause fission products to be released into the coolant circuit. Among the various defects that can form during welding, a root defect at the end plug weld joint is the most prominent type observed in fuel pins. A conventional X-ray film radiographic procedure for end plug welds of fuel pins inserted in a stainless steel (SS) shape correction block with a thickness of 7 mm will limit the minimum detectable wall thickness difference to 1.1% of the specimen thickness due to film gradient and film granularity factors. In the present work, with a standard X-ray digital radiographic data acquisition procedure and optimised radiographic exposure conditions (kV, mAs, etc), the normalised signal-to-noise ratio (SNRn) achieved is 140, which limits the detectable wall thickness difference to 2.8% of the specimen thickness. Hence, a modified digital X-ray radiographic inspection procedure is proposed to assess the quality of end plug welded fuel pins using a suitable X-ray detector calibration technique. The structural noise and pixel intensity variations between the flat-panel pixel elements are minimised using this modified procedure, which results in an order of magnitude increase in the SNRn of the radiographic images. Following implementation of this procedure with end plug welds, the X-ray images show that the achievable wall thickness difference is enhanced to 0.9% of the specimen thickness and microdefects are clearly observed in the welds.

Transient analysis of TMSR-SF0 simulator

ABSTRACT Transient analysis of TMSR-SF0, a molten salt thermal-hydraulic experimental system of molten salt reactor, is investigated using RELAP5/MOD4.0 code. Simulation model consists of two coolant circuits. Variations in core power, primary mass flow, secondary mass flow, and air mass flow are studied separately. Both temporary and permanent phenomena of test facility are concerned in different circumstances. The results indicate that protective system should focus on unusual operation conditions in order to make safety measures for preventing the test facility from overheating and overcooling. The temperature distribution of key measuring points shows that both power and mass flow perturbations are acceptable in simulation model, which is the performance of high heat capacity of molten salt. Meanwhile, the ranges of perturbation are shown in this paper.