Calibration of power monitors in molten salt reactors

Abstract

A method and system for calibrating the nuclear instrumentation of a nuclear reactor are disclosed. The system comprises a reactor configured to operate at a low power level and an external heater configured to heat the reactor to its operating temperature and maintain the reactor in a critical state. The reactor maintains its temperature in a steady state by repeatedly reducing the heater output by changing one or more reactive settings and increasing the reactor output, the reduced heater output being compensated for by using the increased output generated by the reactor itself. The nuclear instrumentation is configured to measure the reactor output and compare it to a calculated reactor output. Based on the discrepancy between the calculated and measured output exceeding a threshold, the nuclear instrumentation is appropriately calibrated by adjusting one or more settings.

Description

[Technical Field]

[0001] Cross-reference of related applications This application claims priority under § 119 of U.S. Provisional Application No. 63 / 518,021, filed on 7 August 2023, which is incorporated herein by reference in its entirety.

[0002] The examples described relate generally to systems, apparatus, and techniques for calibrating nuclear instrumentation in molten salt reactors, particularly for calibrating molten salt reactor power measurement systems at low power levels during startup. [Background technology]

[0003] Molten salt reactors (MSRs) offer an approach to nuclear power that utilizes a molten salt mixture as nuclear fuel instead of the conventional solid fuel used in light water reactors. Advantages include efficient fuel utilization and improved safety (partly by replacing water as a coolant with molten salt). In an MSR, nuclear fission occurs within the molten salt composition contained within the reactor vessel. The reactor vessel can carry hazardous materials, including potentially radioactive substances. Therefore, it is desirable to continuously monitor reactor conditions and parameters such as thermal output, mass flow rate, and temperature at various locations within the reactor system and to inform system administrators for safe and stable operation of the MSR.

[0004] The MSR power level is a quantity that must be measured for both safety and operational concerns. Thermal equilibrium procedures can be applied to calibrate the reactor's nuclear instrumentation (NI) to measure gamma and neutron fluxes, ultimately allowing for the determination of the reactor power. However, when the reactor is in low-power operation before the full-power operating regime is tested, the accuracy of these measurements is limited, leading to measurement results with significant uncertainty. Therefore, low-power calibration of the NI is necessary. [Overview of the Initiative]

[0005] In one example, a method for calibrating the nuclear instrumentation of a nuclear reactor is disclosed. The method teaches that the reactor is operated at a low power level and that a heater is used to heat the reactor to its operating temperature. One or more sensors are used to measure one or more parameters of the reactor. The reactor is configured to maintain its temperature in a steady state by repeating the steps of decreasing the heater power by a predetermined amount in each step and increasing the reactor power by changing one or more reactivity settings until the heater power is zero, in each step the decrease in heater power and increase in reactor power are adjusted to compensate for the decrease in power from the heater by using the increased power generated by the reactor, and the reactor reaches thermal equilibrium.

[0006] The method further teaches that the reactor output is calculated based on measured parameters. Reactor instrumentation is used to measure the reactor output and compare it to the calculated reactor output. If the difference between the calculated and measured reactor output exceeds a threshold, the nuclear instrumentation is appropriately calibrated by adjusting one or more settings.

[0007] In another example, the reactor is a molten salt reactor.

[0008] In another example, the heater is an external energy source.

[0009] In another example, the reactor increases its output while the heater decreases its output.

[0010] In another example, the responsiveness settings include the control rod position, pump output, and flow rate.

[0011] In another example, the nuclear instrumentation comprises two uncompensated ion chambers, two compensated ion chambers, and two fission chambers.

[0012] In another example, nuclear instrumentation is configured to measure neutron flux or neutron count rate and to measure the reactor's output based on the measurement of neutron flux or neutron count rate.

[0013] In another example, the output of a nuclear reactor is calculated based on the energy input from the fuel salt pump, the energy input from the coolant salt pump, the energy input from the heater, the total energy released by one or more sensors, the thermal energy carried by the gas and fluids, the thermal energy carried by the gas and fluids, and the thermal energy carried through the concrete casing of the reactor.

[0014] In another example, one or more parameters measured by one or more sensors include temperature information, energy input from fuel salt pumps, energy input from coolant salt pumps, energy input from heaters, total energy released by one or more sensors, thermal energy carried by gases and fluids, thermal energy carried by gases and fluids, and thermal energy carried through the concrete casing of the reactor.

[0015] In another example, a system for calibrating the nuclear instrumentation of a nuclear reactor is disclosed. The system comprises a reactor configured to operate at a low power level, nuclear instrumentation configured to measure the power of the reactor, a heater configured to heat the reactor to its operating temperature, and one or more sensors mounted via the reactor and configured to measure one or more parameters of the reactor. The reactor maintains its temperature in a steady state by repeatedly taking the steps of decreasing the heater output by a predetermined amount in each step and increasing the reactor output by changing one or more reactivity settings until the heater output is zero, in each step the decrease in heater output and increase in reactor output are adjusted to compensate for the decrease in output from the heater by using the increased output generated by the reactor, and the reactor reaches thermal equilibrium.

[0016] In another example, the system is configured to decrease the output of the heater and simultaneously increase the output of the nuclear reactor.

[0017] In addition to the exemplary aspects described above, further aspects and examples will become apparent by reference to the drawings and consideration of the following description. Brief Description of the Drawings

[0018] [Figure 1A] A schematic block diagram of an exemplary nuclear reactor system is shown. [Figure 1B] A schematic block diagram of a primary heat removal system in an exemplary nuclear reactor system is shown. [Figure 2] A schematic block diagram of an exemplary nuclear reactor system having a nuclear reactor thermal management system is shown. [Figure 3] A schematic block diagram of an exemplary nuclear reactor thermal management system in an exemplary nuclear reactor system is shown. [Figure 4] An exemplary output calibration process for an exemplary nuclear reactor thermal management system in an exemplary nuclear reactor system is shown. [Figure 5] A flow diagram of an exemplary calibration process at a low power level is shown. [Figure 6] A functional block diagram of a computing system is shown. Modes for Carrying Out the Invention

[0019] The use of cross-hatching or shading in the accompanying drawings is generally provided to clarify boundaries between adjacent elements and to facilitate readability of the drawings. Accordingly, the presence or absence of cross-hatching or shading does not convey or indicate any preference or requirement for a particular material, material property, proportion of elements, dimension of elements, commonality of similarly illustrated elements, or any other characteristic, attribute, or property of any element illustrated in the accompanying drawings.

[0020] Furthermore, the (relative or absolute) proportions and dimensions of various features and elements (as well as sets and groups thereof), and the boundaries, separations, and positional relationships presented between them, are provided in the accompanying drawings to facilitate understanding of the various embodiments described herein, and should therefore be understood that they are not necessarily presented or illustrated to scale, and are not intended to indicate any preference or requirement for the illustrated embodiments by exclusion of the embodiments described with reference thereto.

[0021] The following description includes sample systems, methods, and apparatus that embody various elements of this disclosure. However, it should be understood that the disclosure described herein may be implemented in various forms in addition to those described herein.

[0022] The following disclosures relate, in general, to devices, systems, and methods for calibrating nuclear instrumentation (NI) in molten salt reactors (MSRs). NI refers to sensors, detectors, and measuring devices used in reactors and other nuclear facilities to monitor, control, and ensure the safe and efficient operation of the reactor system. These devices can be used to measure and monitor various physical parameters and conditions within the reactor.

[0023] This disclosure describes how a reactor output can be measured by gamma and neutron flux measurements using a NI including a fission chamber, a compensated ion chamber, an uncompensated ion chamber, and / or other instrumentation, and how the measurements can be provided as output values ​​based on calibration related to the core geometry, flux distribution, and the relationship between flux levels and output levels. Based on internal calculations of the instrumentation and known properties of the geometry, the primary value needed to adjust these output values ​​to reflect an actual reactor is an empirical measurement of the system's thermal output, which can be input to the NI via electronic control to properly calibrate the system.

[0024] One method for determining the thermal output for calibrating a nuclear reactor is to determine the thermal equilibrium of the reactor. The following disclosure describes a thermal equilibrium procedure for MSR. Analysis of the thermal equilibrium equation shows that the primary quantities in question are mass flow rate and temperature, and that it is possible to perform calculations based on the thermal equilibrium of air or salt.

[0025] The MSR could be a 1 MWth research reactor. The MSR fuel can use lithium fluoride (LiF) and beryllium fluoride (BeF) salts, commonly known as lithium beryllium fluoride (FLiBe) with a small amount of enriched UF4, and the coolant loop that carries thermal energy from the heat exchanger to the radiator also uses FLiBe. Due to the higher corrosion risk and temperature conditions designed to optimally utilize these materials, relevant sensors with a high level of technical readiness (TRL) for nuclear applications cannot be used in the coolant loop. Similarly, a lower TRL for relevant sensors to directly measure specific FLiBe properties means that such sensors must be considered experimental, as their reliability cannot be adequately demonstrated before the initial operation of the MSR. Therefore, methods for determining the reactor's thermal output may include measuring the heat released into the air by the primary heat removal system (PHRS) radiator and the auxiliary heat removal system (AHRS).

[0026] Referring to the drawings, for illustrative purposes, Figure 1A shows an exemplary MSR100. The MSR100 implements and includes nuclear instrumentation 160 and can implement any of the functions described herein. As to be understood, the example shown in Figure 1A represents only one example configuration of the MSR100 in which such nuclear instrumentation may be used. It will be understood that the nuclear instrumentation described herein can be used in substantially any other configuration of a molten salt reactor as intended herein.

[0027] In various embodiments, the MSR 100 can generate thermal output via nuclear fission reactions using uranium-enriched fuel salt (e.g., high-assay low-enriched uranium). In at least one embodiment, the fuel salt composition may be LiF-BeF2-UF4, but other fuel salt compositions may be used as fuel salt in the reactor system 100. The fuel salt in the system 100 is heated to a high temperature (about 600°C or higher) and melts when the system 100 is heated. In some embodiments, the molten salt reactor system 100 includes a reactor vessel 102 in which the nuclear reaction takes place within the molten fuel salt, a fuel salt pump 104 that pumps the molten fuel salt in a primary loop to a heat exchanger 106 so that the molten fuel salt flows through the heat exchanger and then returns to the reactor vessel, and piping between each component. The heat exchanger 106 is coupled to a secondary loop connected to a coolant salt pump 115, an air-cooled radiator 117, and a coolant drain tank 118. The coolant pump 115 circulates the coolant salt to a heat exchanger, where it absorbs the heat generated by the core from the primary loop and then transfers it to the power generation system. In some examples, the heat generated in the core is transferred to a secondary loop via the heat exchanger and finally released into the atmosphere via a radiator 117.

[0028] Referring further to Figure 1A, the MSR100 may also include, but is not limited to, additional components such as a drain tank 108 and a reactor access vessel 110. The drain tank 108 may be configured to store fuel salt when the fuel salt is in the reactor system 100 but in a subcritical state, and also function as a fuel salt storage unit in the event of a loss of power in the system 100. The reactor access vessel may be configured to allow the introduction of small pellets of uranium fluoride (UF4) into the system 100 as needed to bring the reactor to criticality and compensate for the depletion of fissile material.

[0029] In some examples, the MSR100 may include, among other functions, an inert gas system 112 for supplying inert gas to the upper space of the drain tank 108. The inert gas system 112 may, if necessary, release further inert gas from the upper space of the drain tank 108. Thus, the inert gas system 112 can be operated to maintain insufficient pressurized inert gas in the upper space of the drain tank 108 to substantially prevent molten fuel salt from flowing into the drain tank during normal operation. For example, when the upper space of the drain tank 108 is pressurized by the inert gas system 112, the molten salt can generally circulate between the reactor vessel 102 and the heat exchanger 106 without substantially discharging into the drain tank 108. As described herein, the inert gas system 112 may be configured to supply inert gas to the upper spaces of various other components of the MSR100, such as the upper space of the reactor access vessel 110, and to the seals of the reactor pump 104, among other components. When a shutdown event occurs, the inert gas system 112 can stop supplying inert gas to the space above the drain tank 108 and to other components to which the system 112 supplies inert gas.

[0030] Furthermore, the primary loop, including pumps, tanks, and vessels, can be housed within a reactor casing 130 defining a containment volume 140, the reactor casing being made of concrete. Similarly, the secondary loop, including pumps, tanks, and vessels, can be housed within another reactor casing 131 defining a containment volume 150, the casing also being made of concrete. The primary and secondary loops can work together to form a primary heat removal system (PHRS). As described herein, the PHRS can operate to remove heat from the MSR 100. For example, the fuel salts in the primary loop may generally rise in temperature as a result of fission reactions occurring in the reactor vessel 102. Heat from the molten salts can be transferred from the primary loop by the operation of the secondary loop and the coolant salts circulating within it. For example, the coolant salts in the secondary loop can remove heat from the primary loop via a heat exchanger 106. The secondary loop may be at least partially housed within the containment volume 150 and may include a power generation system (not shown in Figure 1A). To maintain an optimal operating temperature and ensure the safe and stable performance of the reactor, an air-cooled radiator 117 is included in the secondary loop to utilize ambient air to dissipate heat from the coolant salt and transfer absorbed thermal energy to the atmosphere. Nuclear instrumentation (NI) 120 and / or other instrumentation, such as neutron detectors, fission chambers, compensating ion chambers, and uncompensating ion chambers, are located within the containment volume 140 to measure the output of the reactor 100 by gamma and neutron flux measurements and provide the output as an output value to a remotely located instrumentation and control (I&C) system. In some examples, the NI 120 may include a neutron detector configured to measure the neutron count rate within the reactor 100 for output measurement. The I&C system may typically include an MSR human-machine interface (HMI) console that can control all reactor operations and provide various indicators and parameters of the reactor. The I&C system may also include reactor protection systems and industrial safety mechanism activation systems (RPS / EFAS), reaction control systems (RCS), area radiation monitoring systems (ARMS), and HMI console and support systems.In this disclosure, the NI120 may communicate gamma and neutron flux measurements as data signals to an I&C system via one or more communication links.

[0031] Furthermore, one or more sensors can be mounted at different locations throughout the reactor housing 130 to perform various measurements of the reactor 100. For example, thermocouples are commonly used to measure temperatures, including air temperature, salt temperature, and concrete temperature. As an example, sensors 113 and 114 mounted in the reactor housing 130 may include thermocouples configured to measure inlet and outlet air temperatures as well as concrete temperature. As another example, mass flow rates, such as molten salt flow rate, can be measured in a molten salt system having an orifice or Venturi element. For example, sensor 119 may be a Venturi flowmeter used in the reactor 100 to measure molten salt flow rate. Some alternative options include thermal anemometers incorporating thermocouples, startup-based flow sensors, ultrasonic flow sensors, and magnetic flow sensors. It is possible to determine the mass flow rate of salt on the primary or secondary loop using multiple pressure sensors in appropriate locations. Note that sensor 119 may be located elsewhere or incorporated into other components within the reactor 100 to provide different measurements.

[0032] Figure 1B shows a cross-sectional view of a specific component of the MSR100. For example, Figure 1B shows a structure 160 including a first adiabatic region 107, which can be called a “high-temperature atmosphere” in that the temperature of this region is high enough to liquefy the salt and keep all the molten salt contained within it in a molten state. A second adiabatic region 105 can be called a “low-temperature atmosphere” in that it generally has a lower temperature than the first adiabatic region 107. Furthermore, a third adiabatic region 103 can also be called a “low-temperature atmosphere” in that it generally has a lower temperature than each of the first adiabatic region 107 and the second adiabatic region 105. In some cases, the reactor thermal management system (RTMS), as described in detail below and shown in Figures 2 and 3, can operate to maintain and / or control the temperature of the third adiabatic region 103, for example, below a temperature that would cause mechanical fragility of the concrete or structural steel contained therein, thereby maintaining and / or controlling it to a temperature suitable for the operation and / or integrity of a particular component of the system.

[0033] In this disclosure, Figure 1B also shows specific elements of one exemplary auxiliary heat removal system (AHRS). In general, an AHRS may include components and systems for removing heat from the reactor and reactor housing, in addition to the use of coolant salts, as described above. For example, as shown in Figure 1B, one embodiment of an AHRS may include taking in cold air from port 135 and removing heat from the reactor through port 137. In some cases, the passage of air between ports 135 and 147 may be passive, and in other cases, active circulation techniques may be deployed. During operation, one or more sensors may be placed at ports 135 and 137 to measure inlet and outlet airflow rates and temperatures and to use the measurements for thermal equilibrium treatment, as will be described in more detail below with reference to Figures 4 and 5. As shown in Figure 1B, to facilitate the foregoing, the RTMS may include an internal shield 123 associated with or contained within the reactor housing 129.

[0034] For illustrative purposes, Figure 1B further shows a graphite moderator 122 within the reactor vessel 102 defining a channel 126 for the molten salt 109. The reactor vessel 102 is also shown with moderator 122 defining a control rod channel 127 for receiving control rods into it. During operation, the molten salt 109 can circulate through the reactor vessel 102 and associated molten salt loops as described herein. Figure 1B shows the amount of molten salt 109 contained in a drain tank 108. The RTMS can be operated to optionally maintain the molten salt 109 in a molten state for a period of time, even though the molten salt 109 is subcritical and held in the drain tank 108. For example, the RTMS can define various adiabatic barriers described herein to maintain the molten salt 109 in a molten state for a period of time. This allows the system to recirculate and / or move the molten salt elsewhere to repair, replace, or operate the reactor with greater efficiency and safety.

[0035] As described herein, the RTMS of this disclosure may be configured to insulate at least a molten salt reactor vessel and another component (e.g., a “second component” such as a drain tank) that is fluidly coupled to the molten salt reactor vessel. For example, as described above, the RTMS is shown as including an internal vessel or internal shield surrounding the molten salt reactor vessel and the drain tank (e.g., the second component). In other examples, the RTMS of this disclosure may be configured to surround, but not limited to, a reactor access vessel and additional components of the molten salt loop, including some or all of the associated piping of the molten salt loop that fluidly couples the components to each other to form a continuous fluid circuit.

[0036] Figure 2 shows another exemplary MSR200. The MSR200 may be substantially similar to the MSR100 described above in relation to Figure 1A. In this regard, the MSR200 is shown in Figure 2 as including the reactor vessel 202, fuel salt pump 204, heat exchanger 206, drain tank 208, reactor access vessel (RAV) 210, inert gas system 212, coolant salt pump 215, air-cooled radiator 217, coolant drain tank 218, sensor 219, nuclear instrumentation (NI) 220, reactor housing 230 and 231, the redundant descriptions of which are omitted for clarity.

[0037] Despite the aforementioned similarities, the MSR200 in Figure 2 is shown as including a reactor thermal management system (RTMS) 216. The RTMS 216 can be implemented as an integrated insulated enclosure ("furnace") encompassing the reactor vessel 202, drain tank 208, and associated piping, as shown in the functional schematic diagram of Figure 2, as shown in more detail in the example of Figure 3.

[0038] The RTMS216 disclosed herein can be used to maintain and / or control the temperature of one or more components of the MSR200 shown in Figure 2. The RTMS216 may also include an AHRS as shown in Figure 1B (omitted here for ease of presentation). The RTMS216 is shown in Figure 3 as including an insulating layer 302, a reactor vessel 304, heaters 306a-d, a drain tank 308, piping 312a-c, an internal shield, or a vessel 320. During operation, the RTMS216 is configured to maintain and / or control the temperature of the reactor vessel 304 and the drain tank 308. The reactor vessel 304 and the drain tank 308 may be substantially similar to the reactor vessels 102 / 202 and drain tanks 108 / 208 described above in relation to Figures 1A and 2, and for clarity, their redundant description is omitted herein. For example, the reactor vessel 304 and the drain tank 308 may be tanks or containers along a molten salt loop in which heated molten salt is discharged from the reactor vessel through piping 312a and recirculated to the reactor vessel 304 in a cooled form through piping 312b and 312c.

[0039] The inner vessel 320 may be made of an insulating metal (including certain stainless steels) capable of withstanding substantially high temperatures, such as temperatures exceeding 600°C. The inner vessel 320 may surround the reactor vessel 304 and the drain tank 308 to define a first insulating region 310 around it. The first insulating region 310 may be a region of the reactor system that is generally maintained at a temperature sufficient to keep the molten salt in the molten salt loop in a molten state. Thus, the temperature of the first insulating region 310 may, in some cases, exceed 600°C. As will be described in more detail herein, the inner vessel 320 may also function as an additional containment barrier and catch pan in which molten salt can be retained in the event of a leak event from any of the salt-containing components held within the first insulating region 310.

[0040] The insulation layer 302 may be connected to and associated with the inner container 320 to facilitate heat retention within the first insulation region 310. The insulation layer 302 may be further configured to establish a barrier between the first insulation region 310 and the external components of the RTMS 216 (concrete trenches, structural steel, etc.), which may generally require lower temperatures for safe operation and for the components to perform their intended functions. In the example in Figure 3, the insulation layer 302 may be formed from a polytetrafluoroethylene material. In other examples, other suitable insulation materials may be used.

[0041] Heaters 306a-306d may be resistance heaters that are thermally coupled to the inner container 320 or otherwise configured to supply thermal energy to the first adiabatic region 310. In this regard, during operation, heaters 306a-306d may be used to control the temperature of the first adiabatic region 310. For example, heaters 306a-d may be used to heat the first adiabatic region 310 so that its temperature drops below a threshold temperature. Thus, heaters 306a-306d may operate to maintain the first adiabatic region 310 at a temperature that allows the molten salt to remain in a molten state. This may be advantageous, for example, if the inner container 320 is used to capture molten salt from a leak event, and heaters 306a-d operate to supply sufficient thermal energy to the first adiabatic region 310 to keep the leaked molten salt in a molten state within the catch pan or bottom of the inner container 320. In addition, by supplying thermal energy to the first adiabatic region 310, the heaters 306a to 306d can further operate to maintain the reactor vessel 304 at a temperature that allows the molten salt to remain in a molten state, enabling the nuclear fission reaction to occur within the reactor vessel 304 even when the reactor itself is in the initial start-up and low-power stages.

[0042] During operation, heaters 306a-306d function as external energy sources to heat the reactor vessel 304 and liquefy the fuel salts during the initial start-up and low-level phases. As the reactor vessel 304 begins to supply more energy and approaches criticality, the RTMS216 slowly reduces the output of heaters 306a-306d until they are shut down as heat sources. In this process, the reactor output is measured using NI and sensors and compared to a calculated value based on the thermal equilibrium procedure. Based on the comparison, NI is appropriately calibrated. Meanwhile, the reactor 200 in Figure 2 is configured to change its reactivity settings so that the reduced output from heaters 306a-306d is compensated for by the increased output from the reactor 200 itself. For example, the reactor 200 can control its own output by controlling rod position, pump output, flow rate, etc. As shown in Figure 4, the sum of the output of the reactor itself (420) (430) and the output of the external heat sources (410) should be constant.

[0043] Method for measuring output

[0044] The primary value of interest for evaluating the output (power generation) of nuclear reactors (NIs) is the total output of the MSR core. While the local output generated by the reactor varies spatially and has a three-dimensional distribution, the total output involved can be determined by integrating across the entire core. The energy generated by nuclear reactions is released as the kinetic energy of fission products and particles produced by the reactions. This generated energy is proportional to the reactor's flux measurements. Neutron and gamma flux are quantities that can be measured by various detectors. Since the geometric shape and material of the MSR can be considered known quantities, peak flux can be determined by using detectors at known locations and comparing them to the normalized flux distribution expected from theoretical models of a given core configuration. Using the flux distribution determined from theoretical models and measurements, it is possible to integrate the expected distribution and determine the total flux generated based on detector flux measurements, and then use this value to determine the output based on the energy per reaction.

[0045] The most advanced detectors used in the reactor core are designed to perform calibrations from measured flux values ​​to the total internal power, and then directly output the power shown to the operator based on those calibrations. These internal calculations are calibrated based on empirical measurements of power at thermal equilibrium to ensure that the calibrations are properly performed to reflect the actual thermal power generated by the reactor, rather than theoretical relationships from design calculations.

[0046] Nuclear instrumentation for power measurement

[0047] The NI is calibrated by applying thermal equilibrium procedures, and the instrumentation is used to measure gamma and neutron flux, ultimately outputting the reactor power. These NI sensors are selected for their reliability and ability to perform self-calibration and fault detection in conjunction with the primary distributed control system (DCS). Accuracy is also important, as uncertainty must be considered when selecting control setpoints. All of these sensors are located outside the reactor thermal management system (RTMS) but within the reactor casing.

[0048] A NI including a fission chamber, a compensated ion chamber, a non-compensated ion chamber, and / or other instrumentation may be used to measure the output of a reactor through gamma and neutron flux measurements. In this disclosure, the NI may include three sets of different sensors used to measure neutron flux or neutron output. These sensors are provided in pairs to provide redundancy, which helps ensure that minimal maintenance is required throughout the lifetime operation of the MSR. Separate instrumentation paths and independent instruments are utilized to prevent a single failure from disabling multiple detectors.

[0049] The first set of instrumentation, and the only set related to safety, may include two uncompensated ion chambers that function as nuclear safety channels. The ion chambers collect all the charges generated by the incident radiation and the gas contained within the ion chamber to generate a direct current, which can accurately measure the overall ionization effect but not the energy levels of individual ionization events. These are preferable for high radiation dose rates because there is no dead time in the Geiger-Müller tubes that would affect accuracy. The nuclear safety channels primarily focus on ensuring that the design conditions of the reactor fuel salt boundary are not exceeded, and that hypothetical accident scenarios do not damage the fuel salt boundary or disturb the core, associated support structures, or the inside of the reactor during these scenarios by triggering an automated reactor trip. This information can also be linked to the output measurements of all output scenarios.

[0050] The second set of instrumentation may include two compensating ion chambers that function as multi-range linear output channels. By compensating within the ion chamber using a reference chamber, the sensor can electronically cancel out the gamma effect, which can greatly contribute to measurements in the intermediate output range. The linear output channel is primarily used to determine the output in full-power operation.

[0051] A third set of instrumentation may include two fission chambers that function as either multi-range or full-range logarithmic output channels. The fission chambers are in the form of ion chambers lined with highly enriched U-235. Neutrons are not directly ionized, but their interaction with the fissile material can convert them into detectable charged particles. Fission fragments from neutrons interacting with the coating cause significantly more ionization than incident gamma particles. This process makes the fission chambers highly sensitive to neutrons, allowing them to operate in relatively high gamma fields compared to conventional ion chambers. The detector (i.e., the fission chambers) can be used in pulse mode, which allows for highly sensitive detection of individual detector reactions; current mode, which operates similarly to conventional ion chambers; and Campbell mode in high-fluence environments where pulses stack and cannot be accurately measured individually because one pulse is recorded before another pulse disappears, although in this regime the difference between gamma detection and neutron detection is as large as possible so that signal fluctuations can be measured, and the function is a neutron-proportional signal. The overall benefit is that the detector can handle a wide range of outputs using different modes. These sensors are primarily used for starting and low-power operation.

[0052] Combined, NI provides output information across the system's full output range, including subcritical multiplication and conceivable fault scenarios. While calibration using thermal equilibrium is intended to be performed under steady-state conditions, NI shows the output as a function of time during operation.

[0053] Calibration for output measurement

[0054] NI sensors are designed to measure flux, but their primary output is reactor power. They convert local flux measurements at known positions to overall power measurements based on initial calibration from a theoretical model that approximates the expected flux distribution in a given operating regime. Adjusting the calibration using empirical data mainly involves inputting the corrected reactor power measured from its thermal effect to the instrument using the outline of the thermal equilibrium procedure below, updating the internal calculation via system control, which then outputs the corrected power to the operator with an appropriate power display.

[0055] Method for determining thermal equilibrium

[0056] Thermal equilibrium can be calculated by the following procedure by creating a closed system around the air in the system pit. W in -W out =0(1) wherein W in is the energy input to the reactor system per second, and W out is the energy output per second. W in can be decomposed as follows. TIFF0007936575000001.tif14145(2) Here, W FP , W CP , and W RTMS are respectively the energy input per second of the fuel salt pump, the energy input per second of the coolant salt pump, and the energy input per second of RTMS, which can be reduced to 0 when a given system is shut down. They can be determined from consumed power or direct measurement. As described above and shown in Figure 3, the RTMS is implemented as an integrated insulated enclosure ("furnace"), and its energy input is mainly provided by heaters 306a to 306d. The power level of the furnace (i.e., heaters 306a to 306d) can be recorded and measured by the I&C system. W miscThis is the total energy emitted by various minor systems, including but not limited to NI and other sensors operating within the system pit, and can be determined from the consumed output; it should generally be negligible. TIFF0007936575000002.tif12149 is a heat output carried by air, particularly via PHRS and AHRS. TIFF0007936575000003.tif13146 is the heat output generated by the reactor and is a value to be determined. Since thermal equilibrium should be calculated in a steady state, the gas management system is in maintenance mode, mainly circulating helium to prevent stagnation in the gas equipment lines, and is not expected to contribute significantly to the system. If necessary, the contribution of this system can be determined using existing sensors to monitor the gas management system and can be calculated through a process similar to that used to determine the influence of air below. Similarly, since the system pit is expected to be above ambient temperature, it can be assumed that no heat is input through the concrete. Any additional fluid or gas interfaces into the enclosure can be dealt with in a similar manner.

[0057] Similarly, W out It can be broken down as follows: TIFF0007936575000004.tif14141(3) During the ceremony, TIFF0007936575000005.tif13139 is a heat output carried by air, particularly via PHRS and AHRS. TIFF0007936575000006.tif12142 is the thermal energy leaving the system through the enclosure per second. The heat transfer coefficient of the enclosure can be estimated using the integral form of Fourier's law, and a conductive heat transfer equation can be generated treating the concrete as a 1D homogeneous material, which is given by: TIFF0007936575000007.tif19145(4) In the formula, Y is the total number of housing surfaces in the system pit, and U i is the conductivity coefficient of heat transfer across a given surface i, and A i ΔT is the area of ​​surface i. bound.i is the temperature difference across the enclosure material, typically concrete at surface i. The temperature difference can be calculated using the following: ΔT bound.i =T SP,i -T AT,i (5) In the formula, ΔT bound.i This refers to the environment on the opposite side of the enclosure surface i, and the surface i T SP,i The closest measured system pit temperature and ambient temperature T AT,i This is the difference. The system's pit temperature can be measured directly or assumed to be equal to the measured runoff temperature of the AHRS. The ambient temperature is determined from the measurements or fixed by assumptions about environmental conditions such as local soil temperature. Meanwhile, the overall heat transfer coefficient can be determined from the following formula. TIFF0007936575000008.tif16140(6) In the formula, K i is the thermal conductivity of the material at surface i, which has a small temperature dependence but can be approximated based on the ambient temperature, s i is the thickness of the material at surface i. Since both the material composition and thickness of the enclosure boundary are known from the design, the overall heat transfer coefficient of each surface is a known quantity based on the thickness of the material reaching the ambient temperature.

[0058] Furthermore, the details of the heat being carried in and out by the air can be calculated using the following formula. TIFF0007936575000009.tif20141(7) In the formula, X is the total number of air inlet and outlet points in the system, primarily defined by the penetrations as part of the PHRS and AHRS. TIFF0007936575000010.tif11151 is the mass flow rate of air entering the system pit at a given penetration i, and c i This is the specific heat of air in the penetration i, and Ti is the temperature at penetration i. During operation, the mass flow rate of air can be measured by a hot anemometer (with a thermocouple), a Pitot tube, and a vane anemometer. For salt flow rate measurement, a Venturi flow meter, a hot anemometer (with a thermocouple), an activated-base flow sensor (using a CZT or HPGe detector), an ultrasonic flow sensor, and a magnetic flow sensor are used to measure the molten salt (fluid) being introduced (e.g., TIFF0007936575000011.tif12140) and export (for example, The thermal output (TIFF0007936575000012.tif12139) can be measured. Similar detectors are available for other potentially relevant fluids and gases.

[0059] Using these formulas, the following is determined. TIFF0007936575000013.tif21145(8)

[0060] Given that the power inputs to the pumps, furnace (i.e., RTMS), sensors, and various other systems are known or should be easily measured, that the thermal conductivity of the materials, the thickness of the materials, and the surface area are fixed in the structure, and that the absence of pressurized air means that the specific heat of the air changes mainly with temperature, the parameters that must be empirically measured, along with standard assumptions of ambient temperature of the reactor compartment or soil, are the mass flow rate and temperature of the air at each penetration, with the AHRS output temperature corresponding to the pit temperature of the system. Any additional penetrations of the reactor housing require similar measurements.

[0061] Once the thermal properties are empirically measured and the thermal output is derived, this can be compared to the NI output. These sensors are calibrated to determine the bundle at known locations and to give output in units of output. The calculation results can be used to verify that the expected values ​​match a given level of uncertainty, or to adjust them to match thermal equilibrium.

[0062] If a closed system is formed around the salt, similar calculations can be performed, the main difference being that the heat change due to the air is primarily due to PHRS and convection along the piping instead of PHRS and AHRS, and the reactor casing is not part of the boundary of the closed system. Ideally, this would also take into account heat loss from various sensors to air that does not enter the salt. Alternatively, since the heat transferred from the salt matches the heat transferred to the air at that location, salt measurements can be simply used in PHRS calculations if there is less uncertainty.

[0063] The values ​​determined by this method are compared to the power output determined from fuel burnup analysis to validate the measurements. Multiple fuel samples are taken between periods when power measurements are taken, while the reactor is maintained in a steady state. Accurate analysis of various fission products, decay products, and the amount of fuel present makes it possible to determine the power output the reactor was producing from burnup. This value can be compared to both the power output and thermal equilibrium as a cross-check proposed by NI, providing an additional layer of redundancy.

[0064] One problem this method faces is that while the theoretical approach remains valid throughout the entire operating regime, the precision limits of measurements at lower power levels where temperature changes are expected to be small can lead to results with significant uncertainty. Since the MSR is experimented with before the full power operating regime is tested, it is expected to remain at low power for at least several months, and possibly up to a year, during initial startup. In this region, typically planned to reach power up to 10 kW, measurements of temperature changes in the AHRS's 4000 cfm circulation are expected to be negligible, and often appear similar at both the inlet and outlet when uncertainty is considered, even if all the thermal energy expected to be removed by the AHRS system is removed despite the PHRS being inactive at low power. This means that, in practice, this method is unreliable at these low power levels, but it is expected that incremental recalibration of the detector will occur as power increases to reduce uncertainty.

[0065] The most promising method for measuring reactor power at a low level is to perform specific tests using controlled positive reactive insertions, rather than attempting to monitor calibration data online while conducting other studies, when empirical confirmation is deemed necessary in its operating regime. The reactor is set to a known steady state, and all parameters are kept constant while positive reactive insertions are performed. Depending on the specific method chosen, the temperature is kept constant by adjusting the heat input from the furnace, or the furnace power is kept constant while the fuel temperature reaches a new equilibrium. Regardless of which method is chosen, known reactive insertions result in known changes to the expected theoretical models of the reactor's flux distribution and thermal power, which can be taken into account. This calculated change in flux, and the measured NI, can then be compared to the thermal power determined from either the change in power consumed by the furnace reported by the I&C system, or the change in fuel temperature measured using thermocouples placed on the fuel loop for other purposes, which can be used together with the specific heat equation of the fuel salts to determine the change in thermal power. The problem with both of these methods is that they are more sensitive to the small heat loss through penetrations into the fuel loop, which is not a factor in more common techniques that treat the system pit as a black box and analyze the airflow inside and outside the entire system. Furthermore, the presence of fission products requires estimating the specific heat of the fuel salts, as it alters relevant parameters during the course of reactor operation. Finally, all of these experiments are expected to take several hours to several days to reach a new steady state, consuming time that could be used for other experiments. However, these methods are the most accurate techniques available at this time for empirical measurements at low power levels.

[0066] The simplest and currently recommended version of this final technique for performing low-power calibration of nuclear instrumentation is to move the reactor from a steady state in a subcritical configuration heated by a furnace to a steady-state critical configuration achieving the same temperature by shutting down the furnace. The reactor is configured to operate at startup or low power levels. The furnace is used as an external heat source to supply enough energy to liquefy the fuel, e.g., molten salt. The calibration procedure is initiated by cooling the AHRS and heating the subcritical system with an RTMS heater for a sufficient amount of time to reach a steady state, to the following equation: T cal =C(9) TIFF0007936575000014.tif16143(10) In the formula, T cal This is the temperature at which the system is calibrated, and is selected for a given thermocouple. TIFF0007936575000015.tif17143 is a temperature change over time, where C is a constant value dt operator, maintaining the temperature of the thermocouple used for calibration. Ideally, multiple thermocouples should be cross-checked, but the system will check all sensors. The system should remain in a steady state at TIFF0007936575000016.tif12132, but different locations have different constants. The process dt to reach this steady state takes several days. Once this is achieved, the temperature is recorded, and then the reactor becomes critical. The next step is to gradually reduce the heater output, insert reactivity to compensate for the changes, and maintain the temperature in a steady state. During this process, the combination of the output generated by the reactor itself and the output from the external heat source should remain constant, as shown in Figure 3. Ultimately, the heater was shut down and eliminated as the source of both heat and uncertainty, and the reactor was... The reactor is brought back to a steady state as TIFF0007936575000017.tif15144, where dt is expected to take at least several hours, or at least several days. Once the reactor, RTMS, and cells reach a constant temperature again, the reactor power is found to match the initial heater power. It should be noted that the change in power distribution when comparing the furnace and the core means that the temperature of the RTMS and cells should be the focus of such studies, as the temperature distribution of the fuel loop itself may change due to the location of the heat source, provided that sufficient heating is achieved to observe a significant temperature rise above the ambient temperature. The RTMS and cells should maintain a constant temperature C, but the reactor temperature should be kept as close to the initial value as possible during reactivity adjustment. This calibration process is sufficient only at low power, While heater output and reactor output can be matched by setting TIFF0007936575000018.tif14144, the predicted dt performed using this process allows the reactor to be made to higher output, making recalibration using more general procedures applicable, reducing the difficulty of measuring absolute heat lost through AHRS and concrete, and having a relatively low impact on these values. Once operating output is achieved, the output obtained from general procedures should be used for future calibration of the nuclear instrumentation.

[0067] Use thermal equilibrium results to perform calibration of nuclear instrumentation.

[0068] The thermal power of the core determined by the above-described thermal equilibrium Q-core can be used to calibrate the NI. While fission chambers, uncompensated ion chambers, and ion chambers equipped with NI are designed to measure the flux at a given location within the reactor, the power measurements are based on an initial calibration related to a theoretically normalized flux profile and expected peak flux. The overall power of the core is calculated by integrating the expected distribution, and after appropriately scaling the flux using the measurements, the figure of interest is determined using a proportional relationship to the power. Online calibration ensures that the power reflects the actual power of the core when the internal calculations are performed. This also allows operators to account for drift, as the core profile shifts over time due to reactions and depletion in the fuel salts.

[0069] Independent analyses using multiple different sensors can be helpful in cross-checking measurements. With sufficient sensors of adequate reliability, as reflected by TRL, it is possible to determine both the thermal equilibrium of air and salt and whether they coincide within the expected uncertainty. This is particularly useful during startup and low-power operation when fuel isotope analysis is difficult to use. Discrepancies between sensor measurements indicate the need for further analysis, but at this point, thermal equilibrium is treated as the primary method for determining thermal power.

[0070] The practical procedure for utilizing the empirically determined thermal output for calibration primarily involves inputting information into the sensors via their electronic controls. Details of the input methods and internal calculations vary by sensor and manufacturer and are outside the scope of this disclosure at this time.

[0071] Figure 5 shows a flow chart of an exemplary calibration process 500. Process 500 begins at 501. In step 510, the reactor starts operating at a low power level to liquefy a high-melting-point fuel (e.g., molten salt). Exemplary melting points are 190°C to 200°C. Some exemplary molten salts include nitrate mixtures such as LiNO3, KNO3, CaNO3, and NaNO3. In this disclosure, the reactor utilizes lithium fluoride (LiF) and beryllium fluoride (BeF) salts, commonly known as lithium beryllium fluoride (FLiBe) with a small amount of enriched UF4. An external heat source, such as an RTMS, starts operating to heat the reactor to a high temperature and liquefy the fuel. As described in Figure 3, the RTMS is implemented as an integrated insulated enclosure ("furnace"), and its energy input is mainly supplied by heaters 306a to 306d.

[0072] In step 520, with the reactor in a subcritical state, the heat source supplies external energy to heat the reactor, raising its temperature to the operating temperature.

[0073] In step 530, the NI is used to measure the neutron flux or neutron count rate within the reactor. For example, the NI may include two fission chambers located outside the RTMS but within the reactor casing. Based on the measurement of the neutron flux or neutron count rate, the NI can further output the reactor power. In this disclosure, the NI can communicate the measurements to an I&C system remotely located for processing via one or more communication links.

[0074] In step 540, the I&C system can use measurements of neutron flux or neutron count rate to determine whether the reactor is critical. If the reactor is critical, process 500 proceeds to step 550. Otherwise, process 500 returns to step 520.

[0075] In step 550, the I&C is configured to reduce the output of an external heat source, such as heaters 306a-306d of the RTMS as shown in Figure 3, by a predetermined amount. Meanwhile, the I&C system can increase its own output by operating the reactor using control rods, flow rate adjustment, or other reactive control methods. The predetermined reduction in output is compensated for by the increased reactor output generated by changing one or more reactive control settings, such as control rod position, pump output, and flow rate, and as a result, the reactor temperature is maintained in a steady state. In the steady state, the temperature change with respect to time is zero, as described in equations (9) and (10), and the reactor temperature is maintained at a constant value.

[0076] In step 560, the NI is used to measure the neutron flux or neutron count rate in the reactor. For example, the NI may include two fission chambers located outside the RTMS but within the reactor casing. Fission chambers are highly sensitive to neutrons and allow them to operate in a relatively high gamma field compared to a normal ion chamber. They can also handle a wide range of power levels using different modes. Therefore, fission chambers are primarily used for startup and low-power operation. Based on the measured neutron flux or neutron count rate, the NI can output the reactor power and communicate the measurement to the I&C system. The I&C system is configured to compare the output of the NI with the thermal power of the reactor, calculated by equation (8). If the difference exceeds the expected level of uncertainty, the NI can be appropriately calibrated. During operation, particularly during startup and low-power operation, the increase in reactor power at each step (i.e., step 550) is determined by tracking the output change of the external heat source while keeping all other variables constant, so that the increased reactor power matches the decreased output of the external heat source, provided that the temperature at which NI is calibrated is stably maintained at a constant value, and by bringing the reactor to thermal equilibrium for each step.

[0077] If, in step 570, the external heat source is at zero output and the reactor is supplying all the heat to maintain its operating temperature, process 500 proceeds to step 580. Otherwise, process 500 proceeds to step 550, where the output of the external heat source continues to decrease and the output generated by the reactor itself continues to increase.

[0078] In step 580, the I&C system is configured to shut down the external heat source (e.g., heaters 306a-d of the RTMS as shown in Figure 3) when its output is zero. The reactor itself can maintain its operating temperature and supply sufficient output to operate in a steady state. The NI can then continue measuring the neutron flux or neutron count rate, output the reactor output, and then calibrate its sensors at a high output level based on a comparison with the thermal output calculated by equation (8). The process then stops in step 599.

[0079] Figure 6 shows a functional block diagram of the computing system 600. The schematic diagram in Figure 6 generally represents any type of system and configuration that may be used to receive and process various signals from NI and sensors as described herein. For example, the computing system 600 may be used with or contained within an NI and I&C system and may be used to perform any of the functions described herein. In this regard, the computing system 600 may include any suitable hardware (e.g., computing devices, data centers, switches), software (e.g., applications, system programs, engines), network components (e.g., communication paths, interfaces, routers), etc. (not necessarily shown for clarity) used to facilitate any suitable operation disclosed herein.

[0080] As shown in Figure 6, the computing system 600 may include a processing unit or element 601 operably connected to computer memory 602 and computer-readable media 603. The processing unit 601 may be operably connected to the components of memory 602 and computer-readable media 603 via an electronic bus or bridge (e.g., a system bus 607). The processing unit 601 may include one or more computer processors or microcontrollers configured to perform operations in response to computer-readable instructions. The processing element 601 may be the central processing unit of the computing system 600. Additionally or alternatively, the processing unit 601 may be other processors in a device, including application-specific integrated circuits (ASICs) and other microcontroller devices.

[0081] Memory 602 may include various types of non-temporary computer-readable storage media, such as read-access memory (RAM), read-only memory (ROM), erasable programmable memory (e.g., EPROM and EEPROM), or flash memory. Memory 602 is configured to store computer-readable instructions, sensor values, and other persistent software elements. Computer-readable media 603 may also include various types of non-temporary computer-readable storage media, such as hard drive storage devices, solid-state storage devices, portable magnetic storage devices, or other similar devices. Computer-readable media 603 may also be configured to store computer-readable instructions, sensor values, and other persistent software elements.

[0082] In this example, the processing unit 601 is operable to read computer-readable instructions stored in memory 602 and / or computer-readable medium 603. The computer-readable instructions can be adapted to the processing unit 601 to perform the operations or functions described above with respect to Figures 1 to 5. The computer-readable instructions may be provided as a computer program product, a software application, etc.

[0083] As shown in Figure 6, the computing system 600 may also include a display 604. The display 604 may include liquid crystal displays (LCDs), organic light-emitting diode (OLED) displays, light-emitting diode (LED) displays, and the like. If the display 604 is an LCD, the display may also include a backlight component that can be controlled to provide a variable level of display brightness. If the display 604 is an OLED or LED type display, the brightness of the display 604 can be controlled by changing the electrical signals supplied to the display elements.

[0084] The computing system 600 may also include a battery configured to supply power to the components of the computing system 600. The battery may include one or more energy storage cells connected to each other to provide an internal supply of power. In this regard, the battery may be a component of the power supply 605 (for example, including a charging system or other circuit that supplies power to the components of the computing system 600). The battery may be operably coupled to a power management circuit configured to provide appropriate voltage and power levels to individual components or groups of components within the computing system 600. The battery may be configured to receive power from an external power source, such as an AC power outlet or an interconnected computing device, via the power management circuit. The battery may store received power for extended periods, sometimes ranging from several hours to several days, so that the computing system 600 can operate without being connected to an external power source.

[0085] The computing system 600 may also include a communication port 606 configured to transmit and / or receive signals or telecommunications from an external or separate device. For example, in this disclosure, the computing system 600 in an I&C system is configured to transmit and / or receive signals or telecommunications representing measurements from sensors in NI and MSR. The communication port 606 may be configured to connect to an external device via a cable, adapter, or other type of electrical connector. In some embodiments, the communication port 606 can be used to connect the computing system 600 to a computing device and / or other suitable accessories configured to transmit and / or receive electrical signals. The communication port 606 may be configured to receive identification information from an external accessory that can be used to determine a mounting or support configuration. For example, the communication port 606 can be used to determine that the computing system 600 is connected to a mounting accessory such as a particular type of stand or support structure.

[0086] Other examples and embodiments are within the scope and spirit of the claims of this disclosure and the appended patents. For example, features implementing a function may also be physically located in various positions, including being distributed so that some of the functions are implemented in different physical locations. The foregoing description uses a specific nomenclature to provide a complete understanding of the examples described for illustrative purposes. However, it will be apparent to those skilled in the art that specific details are not required to implement the examples described. Accordingly, the foregoing description of the specific examples described herein is presented for illustrative and explanatory purposes only. They are not intended to be exhaustive or to limit the embodiments to the exact forms disclosed. Given the above teachings, it will be apparent to those skilled in the art that many modifications and variations are possible.

Claims

1. A method for calibrating the nuclear instrumentation of a nuclear reactor, A step of starting the reactor and heat source, wherein the heat source is a heater, Following the aforementioned starting step, the step of heating the reactor to the operating temperature by using the heater, After the heating step, a step is taken to determine whether the reactor is in a critical state, The steps include:

1. Receiving a determination that the reactor is in a critical state, and repeating the step of reducing the output of the heater by a predetermined amount; A step of increasing the output of the reactor by changing one or more reactivity settings, The steps of reducing the output of the heater and increasing the output of the reactor are adjusted to compensate for the reduced output from the heater until the reactor reaches thermal equilibrium by using the increased output generated by the reactor. After the step of increasing, the step of calculating the output of the reactor based on the heat transfer function of a plurality of parameters of the reactor, After the step of increasing the output, the step of measuring the output of the reactor using the nuclear instrumentation, After the calculation and measurement steps, the nuclear instrumentation is calibrated based on the calculated output of the reactor and the measured output of the reactor. After the calibration step, the step of determining that the output of the heater is zero and the temperature of the reactor is the operating temperature, A method comprising the step of stopping the heater after determining that the output of the heater is zero and the temperature of the reactor is the operating temperature.

2. The method according to claim 1, wherein the reactor is a molten salt reactor.

3. The method according to claim 1, wherein the heater is an external energy source located outside the inner vessel of the reactor.

4. The method according to claim 1, wherein the step of reducing the output of the heater and the step of increasing the output of the reactor are performed simultaneously.

5. The method according to claim 1, wherein the reactivity setting includes the control rod position, pump output, and flow rate.

6. The method according to claim 1, wherein the nuclear instrumentation comprises two uncompensated ion chambers, two compensated ion chambers, and two fission chambers.

7. The aforementioned nuclear instrumentation, By measuring the neutron flux or neutron count rate, The method according to claim 1, configured to measure the output of the reactor based on the measurement of neutron flux or neutron count rate.

8. The method according to claim 1, wherein the output of the reactor is calculated based on the heat transfer function of the plurality of parameters measured by a plurality of sensors installed throughout the reactor, the plurality of parameters include energy input from the fuel salt pump, energy input from the coolant salt pump, energy input from the heater, the sum of the energy released by one or more sensors, thermal energy carried by gas and fluid, thermal energy brought in by gas and fluid, and thermal energy escaping to the outside through the concrete casing of the reactor.

9. The step of heating the reactor using the heater is: The steps include: Temporarily stopping the heater upon determining that the temperature of the reactor has reached a threshold temperature; The method according to claim 1, further comprising the step of restarting the heater after determining that the temperature of the reactor has fallen below the threshold temperature.

10. The method according to claim 1, wherein the repeated step is stopped upon determination that the output of the heater is zero and the temperature of the reactor is the operating temperature.

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