A reactor primary loop heat pipe section average temperature correction calculation method

Abstract

The present application belongs to the field of nuclear power plant reactor operation management and core supervision, and particularly relates to a reactor primary loop heat pipe section average temperature correction calculation method. i The method comprises the following steps: step 1: calculating the reactor power W for the enthalpy rise of the assembly; step 2: calculating the outlet coolant temperature T of the core fuel assembly; step 3: establishing a temperature distribution interval {T k}; step 4: calculating the coolant mass share in each group, i.e. the measuring point weight coefficient; step 5: calculating the average temperature of each temperature grouping interval; step 6: calculating the temperature offset of each measuring point of the heat pipe section; step 7: calculating the heat pipe section weight average temperature for each primary loop; step 8: calculating the loop heat power; and step 9: obtaining the total reactor power. The present application has the beneficial effect that the primary loop heat balance power is evaluated, i.e. the primary loop heat balance power is calculated using the heat pipe section weight average temperature, and compared with the secondary loop heat balance power, which is obviously improved compared with the traditional method.

Description

Technical Field

[0001] This invention belongs to the field of nuclear power plant reactor operation management and core monitoring, and specifically relates to a method for calculating the average temperature correction of the primary loop heat pipe section of a reactor. Background Technology

[0002] Traditional methods for calculating the average temperature of the heat pipe section primarily use the arithmetic mean of multiple measuring points within the heat pipe section as the average temperature. However, the primary loop thermal power calculated using this temperature generally deviates significantly from the secondary loop thermal power. Based on engineering experience, the secondary loop thermal power is generally considered to be accurately calculated. Therefore, the large power deviation is due to inaccurate calculations of the primary loop thermal power. The root cause is the uneven temperature distribution of the coolant in the primary loop heat pipe section, leading to inaccurate measurement of the average temperature. Existing measurement methods are not suitable for current reactor operations, thus requiring modification of the measurement and calculation methods.

[0003] Currently, when nuclear power plants calculate the average temperature of the heat pipe section in the primary loop of the reactor, they generally calculate the arithmetic mean of the temperatures measured by multiple resistance temperature detectors (RTDs) for the same heat pipe section. The arithmetic mean temperature is then used as the average temperature of the coolant in that heat pipe section, and further used for calculations of the thermal power of the reactor primary loop.

[0004] Due to the uneven radial flow distribution and uneven radial power distribution in the reactor core, the temperature rise of the primary coolant after flowing through the core is uneven. When the coolant that has not been fully stirred enters the heat pipe section of the primary main pipe, uneven radial temperature distribution will occur in the pipe. Therefore, the temperature values ​​measured by different thermal resistors at the same cross-sectional position in the heat pipe section will have large deviations.

[0005] If the arithmetic mean of the measured temperature values ​​of different thermal resistors at the same cross-sectional location within the same heat pipe section is directly used as the average temperature of that heat pipe section, it will deviate from the physical reality, resulting in a large deviation in the calculation of the average temperature of the heat pipe section, and consequently, deviation in the calculation of parameters such as core thermal power. Summary of the Invention

[0006] The purpose of this invention is to provide a method for correcting the average temperature of the heat pipe section in the primary loop of a reactor. By correcting the existing traditional method for calculating the average temperature of the heat pipe section, the calculation error of the traditional method is reduced, and the corrected average temperature of the heat pipe section can be closer to the true average temperature.

[0007] The technical solution of the present invention is as follows: A method for calculating the average temperature correction of the primary loop heat pipe section of a reactor, comprising the following steps:

[0008] Step 1: Calculate the reactor power W required for the enthalpy rise of the assembly;

[0009] Step 2: Calculate the coolant temperature T at the fuel assembly outlet of the reactor core.i ;

[0010] Step 3: Establish the temperature distribution range {T} k}:

[0011] Step 4: Calculate the coolant mass share within each group, i.e., the measurement point weighting coefficient;

[0012] Step 5: Calculate the average temperature for each temperature group interval;

[0013] Step 6: Calculate the temperature offset at each measuring point in the heat pipe section;

[0014] Step 7: For each primary loop, calculate the weighted average temperature of the heat pipe segment;

[0015] Step 8: Calculate the loop heat power;

[0016] Step 9: Obtain the total power of the reactor.

[0017] Step 1 obtains the reactor secondary loop thermal power, main pump power, heat loss power, and pressurizer heating power. The secondary loop thermal power is subtracted from the main pump power, and the heat loss power is added. Then, the pressurizer heating power is subtracted to obtain the reactor power W used to calculate the enthalpy rise of the assembly.

[0018] Step 2 is to obtain the relative power distribution H of the core fuel assemblies. i Relative flow distribution of fuel assemblies in the reactor core F i Average temperature of cold pipe section The core fuel assembly outlet coolant temperature T is calculated using the following formula. i :

[0019]

[0020] Step 3 involves obtaining the temperature value T at each measuring point in each loop heat pipe section. k Establish the temperature distribution interval {T} according to the following rules. k}:

[0021]

[0022] Step 4 involves adjusting the coolant temperature T at the core fuel assembly outlet. i Group the coolant according to the distribution intervals in step 3, and obtain the total primary coolant flow rate F. pri To obtain the distribution share η of coolant in each loop within the fuel assembly. i Meanwhile, the bypass flow share is equivalent to an additional fuel assembly whose outlet coolant temperature is equal to the average temperature of the cold pipe section. The coolant mass share within each group, i.e., the measurement point weighting coefficient, is calculated using the following formula:

[0023]

[0024]

[0025] Step 5 involves calculating the average temperature for each temperature group interval using the following formula:

[0026]

[0027] Step 6 involves calculating the temperature offset at each measuring point in the heat pipe section using the following formula:

[0028]

[0029] Step 7 involves calculating the weighted average temperature of the heat pipe segment for each primary loop:

[0030]

[0031] Step 8 involves obtaining the total coolant flow rate F in the loop. loop Use the water property parameter table to look up the enthalpy value of the loop cooling pipe section. and enthalpy of heat pipe section For each primary loop heat, calculate the loop heat power:

[0032]

[0033] Step 9 involves summing the power of each loop to obtain the total reactor power.

[0034] W pri =∑W loop .

[0035] The beneficial effects of this invention are as follows: it evaluates the primary loop heat balance power, that is, it uses the weighted average temperature of the heat pipe section to calculate the primary loop heat balance power and compares it with the secondary loop heat balance power. If the power deviation is significantly improved compared with the traditional method, it indicates that the modified calculation method is effective.

[0036] Taking a nuclear power plant as the subject, periodic test data parameters for more than 20 cycles were collected from multiple units. The core conditions of the collected data were all rated operating conditions, including core assembly power distribution, reactor secondary loop thermal balance power (including main pump power, pressurizer power, heat loss power, etc.), temperature values ​​of each measuring point in the primary loop hot and cold pipe sections, etc. Assembly flow distribution and mass share entering each loop were based on publicly available data.

[0037] Calculations have verified that, for the verified cycle, if the arithmetic mean temperature of the heat pipe section of the primary loop is calculated using the traditional method, the deviation of the heat balance power between the primary and secondary loops is generally above 2%, especially for VVER units, where the deviation can reach 3%. However, when the weighted average temperature of the heat pipe section of the primary loop is calculated using this method, the deviation of the heat balance power between the primary and secondary loops can be kept within 1%.

[0038] By using a pre-made data table of measurement point weight coefficients and temperature offsets according to a certain fuel consumption interval, the weighted average temperature of the heat pipe section of the primary loop is calculated for the balanced cycle. At this time, the deviation of the calculated primary and secondary loop thermal balance power can also be kept within 1%.

[0039] This method is applicable in principle to all types of pressurized water reactor nuclear power units. However, considering the differences in reactor measurement systems, such as the fact that VVER units have more temperature measurement probes in the primary loop heat pipe section (6), and that the primary loop thermal power of VVER units is an important reference for the displayed value of reactor thermal power instruments, the effect of this method is relatively more significant when applied to VVER units. Attached Figure Description

[0040] Figure 1 This is a schematic diagram illustrating the principle of a method for calculating the average temperature correction of the primary loop heat pipe section of a reactor, as provided by the present invention. Detailed Implementation

[0041] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments.

[0042] Based on the thermal characteristics of the reactor and the primary loop, this invention provides a method for calculating the average temperature correction of the primary loop heat pipe section. This method specifically corrects the uneven temperature distribution of the primary loop coolant within the heat pipe section, thereby calculating the corrected average temperature of the primary loop heat pipe section and reducing the average temperature deviation of the primary loop heat pipe section.

[0043] The method for correcting the average temperature of the heat pipe section in the primary coolant loop of a reactor can correct the average temperature of the heat pipe section based on the temperature field distribution of the primary coolant in the primary coolant system. The corrected calculation model considers the non-uniformity of coolant heating in the reactor core, including non-uniform radial power distribution and non-uniform radial flow distribution in the core. In addition, it also needs to consider the bypass flow rate of the primary coolant in the core.

[0044] like Figure 1As shown, after the primary coolant enters the reactor through the reactor pressure vessel inlet via the loop cold pipe section, most of the coolant flows downwards along the pressure vessel annulus into the lower core chamber and then upwards. After flow distribution, it flows upwards into the fuel assemblies, with the main flow direction being upwards along the fuel assembly axis. After being heated by the assemblies or separated by the core shroud, it enters the upper core chamber. In addition, a very small portion of the coolant flowing into the reactor through the pressure vessel inlet flows upwards to cool structural components such as the reactor top cover, and then merges into the upper core chamber. All the coolant mixes in the upper chamber and then enters the primary heat pipe section through the reactor pressure vessel outlet.

[0045] To measure the temperature of the reactor's primary coolant circuit, several resistance temperature detectors (RTDs) are installed near the reactor outlet in the hot pipe section and near the reactor inlet in the cold pipe section. These RTDs are typically arranged uniformly along the axial direction within the same cross-section (e.g., at 120-degree intervals). For the cold pipe section, since the coolant is sufficiently cooled, the measured temperature values ​​at all measuring points within the same loop are essentially consistent. However, for the hot pipe section, for most reactors, varying degrees of temperature measurement deviations have been observed at different measuring points within the same loop's hot pipe section.

[0046] The current common method for calculating the average coolant temperature of hot or cold pipe sections is to directly calculate the arithmetic mean of the measurements taken at all measuring points within the same loop. Based on statistical principles, multiple measurements under the same conditions can reduce random errors; in this case, the arithmetic mean of the valid measurements can be used as the measurement result. The observed temperature measurement deviations at different measuring points in the loop's hot section can exclude deviations caused by random factors. These deviations reflect a real physical phenomenon: the uneven temperature distribution within the loop's hot pipe section. Therefore, from a thermodynamic perspective, using the arithmetic mean of temperatures measured at different measuring points in a fluid with uneven temperature distribution as its average temperature assumes that the mass fraction of the fluid characterized by the measuring point temperature is the same. However, in actual physical fields, if the mass fraction of the fluid varies with temperature distribution, using the arithmetic mean of temperatures at different measuring points as the average temperature of the entire flow field may lead to significant deviations.

[0047] When measuring the average temperature of a flow field with temperature distribution, if the arithmetic mean of different measuring points is used, based on the principle of differentiation, a relatively close measurement value can only be obtained when the number of measuring points reaches a certain level. However, in reality, it is impossible to arrange a sufficient number of measuring points in the main piping of the reactor primary loop to meet the measurement accuracy requirements. Therefore, with a limited number of measuring points, calculating the average temperature of the heat pipe section of the main piping of the reactor primary loop with temperature distribution requires consideration of correcting the fluid fraction, i.e., the weighting coefficient correction method.

[0048] As above Figure 1As shown, in reactor design, considering factors such as core power distribution, the coolant flow rate entering each fuel assembly in the core exhibits a certain degree of non-uniformity. For assembly A, the relative coolant flow rate in the mainstream direction is F1, and the relative power in a certain state is H1; for assembly B, the relative coolant flow rate in the mainstream direction is F2, and the relative power in the same state is H2. Therefore, the relative enthalpy rises of assemblies A and B are respectively:

[0049]

[0050] The outlet temperatures of components A and B are respectively:

[0051]

[0052] If we ignore the nonlinear change in enthalpy with temperature within the fuel assembly outlet coolant temperature range (in reality, the maximum temperature difference between the fuel assembly outlet coolants is generally around 20°C, and the error introduced by this linear assumption is very small), then the average temperature of the outlet coolants for assemblies A and B is:

[0053]

[0054] For a reactor core with N fuel assemblies, the average temperature of the coolant at the assembly outlet is:

[0055]

[0056] In other words, if a temperature sensor is installed at each fuel assembly outlet to measure the temperature of the coolant after heating at each fuel assembly outlet, let's assume the measured temperature is:

[0057]

[0058] The fuel assembly outlet temperatures are divided into several groups according to the temperature gradient (e.g., group K), and the representative temperature of each group is taken as T from low to high. k (k = 1, 2, 3... K), where the first temperature range is denoted as {T1}, and the rule is:

[0059]

[0060] The temperature range of group k is (1 < k < K), and the temperature range of group k is denoted as {T}. k The rules are as follows:

[0061]

[0062] The temperature range of group K is denoted as {T} K The rules are as follows:

[0063]

[0064] Therefore, the average temperature of the k-th group is:

[0065]

[0066] Based on the above total of K representative temperature values, the average core outlet coolant temperature is:

[0067]

[0068]

[0069]

[0070]

[0071] After being mixed in the upper chamber of the reactor, the coolant enters the heat pipe section of the primary loop through the reactor outlet. Although the mass fraction of the coolant temperature distribution changes relative to the upper chamber due to energy exchange, based on the principle of energy conservation, and ignoring heat losses or other external heating sources, the overall energy of the system remains constant. It is merely an energy exchange within different groups of the system, and the aforementioned principle for calculating the average temperature still holds true. However, the proportion of coolant flowing through the fuel assemblies entering each loop needs to be considered. Therefore, when grouping the coolant temperature heated by the reactor core according to temperature gradients, the characteristic temperature T of each group... k The measured temperature at each measuring point in the heat pipe section is defined as the average temperature of the heat pipe section calculated using the following formula, which is called the weighted average temperature. ω k Defined as the weighting coefficient of the measuring points in the heat pipe section, ΔT k Defined as the temperature offset at the measuring point, the weighted average temperature of the heat pipe section is:

[0072]

[0073] in:

[0074]

[0075]

[0076]

[0077]

[0078]

[0079] Among them, T k F represents the temperature at each measuring point in the loop heat pipe section. pri F represents the total coolant flow rate in the primary coolant loop, taking into account the core bypass flow rate. iThis refers to the flow share of each component in the reactor core, as well as the bypass flow share; η i The mass fraction of coolant entering a specific loop heat pipe section within the fuel assembly; H i The relative power of each component in the reactor core is given, while the heating power of the bypass flow is 0. is the average temperature of the cold pipe section of the primary loop; W is the reactor thermal power, which is generally calculated from the thermal balance power of the secondary loop. This is a function that utilizes enthalpy rise and calculates coolant temperature based on a temperature-entropy diagram.

[0080] Based on the above principle, the approximate average temperature of the coolant in the heat pipe section of the reactor primary loop can be calculated at any given time.

[0081] Since the core power distribution varies with burnup and operating conditions (including control rod position, power level, etc.), the weighting coefficients and temperature offsets of each measuring point in the primary loop heat pipe section also change. For a given fuel-loaded core, the power distribution at each burnup point under design conditions is known. Ignoring the influence of the power distribution on the core flow distribution, the weighting coefficients and temperature offsets of the measuring points can be compiled into a data table according to a certain burnup interval, which can be used as the initial configuration data for the measurement system.

[0082] This invention is used for processing periodic test data. Compared to traditional methods that often lead to power deviations and other test results exceeding the test acceptance criteria, this method ensures that power deviations and other test results remain within the test criteria range. This method can also be used for the accurate measurement of the average temperature of the primary loop heat pipe section and the reactor thermal power.

[0083] A method for calculating the average temperature correction of the primary loop heat pipe section of a reactor includes the following steps:

[0084] Step 1: Obtain the reactor secondary loop thermal power, main pump power, heat loss power, and pressurizer heating power. Subtract the main pump power, heat loss power, and pressurizer heating power from the secondary loop thermal power to obtain the reactor power W used to calculate the enthalpy rise of the components.

[0085] Step 2: Obtain the relative power distribution H of the reactor core fuel assemblies i Relative flow distribution of fuel assemblies in the reactor core F i Average temperature of cold pipe section The core fuel assembly outlet coolant temperature T is calculated using the following formula. i :

[0086]

[0087] T is the input of the pressure and enthalpy of unsaturated water, and the water temperature is looked up from the water property parameter table. In this patent, the standard for thermodynamic properties published by the International Association for the Study of Water and Steam Properties (IAPWS-IF97) in 1997 is referenced for lookup and calculation. The input pressure is the pressure P at the top of the reactor core. out The input enthalpy value is the component outlet enthalpy H. out The pressure P at the top of the reactor core out Output from the core measurement system, H out Calculations based on component inlet temperature and enthalpy rise:

[0088] H out =H in +ΔH

[0089]

[0090] H in The calculation uses the pressure and temperature of unsaturated water as input, and the enthalpy of water is obtained from a table of water properties. This patent references the IAPWS-IF97 standard for thermodynamic properties published by the International Association for the Study of Water and Steam Properties in 1997 for reference and calculation. The input pressure is the core bottom pressure P. in The input temperature is This represents the average temperature of the cold pipe section in the primary loop. The core bottom pressure P... in The average temperature of the primary loop cold pipe section, output from the core measurement system. Output calculated by the reactor core measurement system.

[0091] Step 3: Obtain the temperature value T at each measuring point in each loop heat pipe section. k Establish the temperature distribution interval {T} according to the following rules. k}:

[0092]

[0093] K is the total number of groups, k is the group number, T1 is the temperature of measuring point 1, and T2 is the temperature of measuring point 2.

[0094] Step 4: Set the core fuel assembly outlet coolant temperature T i Group the coolant according to the distribution intervals in step 3, and obtain the total primary coolant flow rate F. pri To obtain the distribution share η of coolant in each loop within the fuel assembly. i Meanwhile, the bypass flow share is equivalent to an additional fuel assembly whose outlet coolant temperature is equal to the average temperature of the cold pipe section. The coolant mass share within each group, i.e., the measurement point weighting coefficient, is calculated using the following formula:

[0095]

[0096]

[0097] gi is a logical function that has a value of 1 if a certain condition is met, and 0 otherwise.

[0098] Step 5: Calculate the average temperature for each temperature group interval using the following formula:

[0099]

[0100] Step 6: Calculate the temperature offset at each measuring point of the heat pipe section using the following formula:

[0101]

[0102] Step 7: For each primary loop, calculate the weighted average temperature of the heat pipe segment:

[0103]

[0104] Step 8: Obtain the total coolant flow rate F in the loop. loop Use the water property parameter table to look up the enthalpy value of the loop cooling pipe section. and enthalpy of heat pipe section For each primary loop heat, calculate the loop heat power:

[0105]

[0106] Step 9: Add up the power of each loop to obtain the total reactor power:

[0107] W pri =∑W loop

[0108] This invention can also be implemented using pre-made tables. Since the core power distribution changes with burnup and operating conditions (including control rod position, power level, etc.), the weighting coefficients and temperature offsets of each measuring point in the primary loop heat pipe section also change. For a given fuel-loaded core, the power distribution at each burnup point under design conditions is known. Ignoring the influence of the power distribution on the core flow distribution, the weighting coefficients and temperature offsets of the measuring points can be compiled into a data table according to a certain burnup interval. This table serves as the initial configuration data for the measurement system, which can then be read and used by the measurement system program.

[0109] The calculation logic for the weighting coefficients and temperature offset is as follows:

[0110] 1) Based on the reactor's rated power state W, rated rod position, and rated coolant operating flow rate F. pri Coolant operating temperature of cold pipe section Based on this, calculate the core assembly power distribution H at each burnup point. i ;

[0111] 2) Obtain the relative flow distribution F of the core fuel assembly i The core fuel assembly outlet coolant temperature T is calculated using the following formula. i :

[0112]

[0113] 3) Obtain the temperature values ​​T at each measuring point of each loop heat pipe section operating under rated conditions at the same fuel consumption point under historical cycle (balanced cycle) conditions. k Establish the temperature distribution interval {T} according to the following rules. k}:

[0114]

[0115] 4) Set the core fuel assembly outlet coolant temperature T i Group the components according to the distribution intervals in step 3, and obtain the distribution share η of the coolant in each loop within the fuel assembly. i Meanwhile, the bypass flow share is equivalent to an additional fuel assembly whose outlet coolant temperature is equal to the average temperature of the cold pipe section. The coolant mass share within each group, i.e., the measurement point weighting coefficient, is calculated using the following formula:

[0116]

[0117]

[0118] 5) Calculate the average temperature for each temperature group interval using the following formula:

[0119]

[0120] 6) Calculate the temperature deviation at each measuring point in the heat pipe section using the following formula:

[0121]

[0122] The table style can be created as follows:

[0123] Table 1. Weighting coefficients and temperature offset parameter configurations for each measuring point in each loop heat pipe section of the reactor primary loop (where: superscript indicates loop number, subscript indicates measuring point number).

[0124]

[0125] Based on the table above, during actual reactor operation, the weighting coefficients and temperature offset parameter configuration data of each measuring point in the heat pipe section of each loop in the reactor primary loop under similar burnup conditions can be read to calculate the weighted average temperature of each heat pipe section in the reactor primary loop:

[0126] 7) For each primary loop, obtain the current fuel consumption value TEF and the temperature value T at each measuring point in each heat pipe section. k Based on the combustion point and the temperature value of the measuring point, a weighting coefficient ω is matched. k and temperature offset ΔT k ;

[0127] 8) For each primary loop, calculate the weighted average temperature of the heat pipe section:

[0128]

[0129] 9) Obtain the total coolant flow rate F in the loop. loop Use the water property parameter table to look up the enthalpy value of the loop cooling pipe section. and enthalpy of heat pipe section For each primary loop heat, calculate the loop heat power:

[0130]

[0131] 10) Add the power of each loop to obtain the total reactor power:

[0132] W pri =∑W loop .

Claims

1. A method for calculating and correcting the average temperature of the heat pipe section in the primary loop of a reactor, characterized in that, Includes the following steps: Step 1: Calculate the reactor power required for component enthalpy rise ; Step 2: Calculate the coolant temperature at the fuel assembly outlet of the reactor core. ; Step 3: Establish temperature distribution range : Step 4: Calculate the coolant mass share within each group, i.e., the measurement point weighting coefficient; Step 4 involves adjusting the coolant temperature at the core fuel assembly outlet. Group the coolant according to the distribution intervals in step 3, and obtain the total primary coolant flow rate. To obtain the distribution of coolant in each loop within the fuel assembly. Meanwhile, the bypass flow share is equivalent to an additional fuel assembly whose outlet coolant temperature is equal to the average temperature of the cold pipe section. The coolant mass share within each group, i.e., the measurement point weighting coefficient, is calculated using the following formula: ， ， Step 5: Calculate the average temperature for each temperature group interval; Step 5 involves calculating the average temperature for each temperature group interval using the following formula: ， Step 6: Calculate the temperature offset at each measuring point in the heat pipe section; Step 7: For each primary loop, calculate the weighted average temperature of the heat pipe segment; Step 8: Calculate the loop heat power; Step 9: Obtain the total power of the reactor.

2. The method for calculating the average temperature correction of the primary loop heat pipe section of a reactor as described in claim 1, comprising the following steps: Step 1 obtains the secondary loop thermal power, main pump power, heat loss power, and pressurizer heating power; subtracts the main pump power from the secondary loop thermal power, adds the heat loss power, and then subtracts the pressurizer heating power to obtain the reactor power used to calculate the enthalpy rise of the assembly. .

3. The method for calculating the average temperature correction of the primary loop heat pipe section of a reactor as described in claim 1, comprising the following steps: Step 2 is to obtain the relative power distribution of the fuel assemblies in the reactor core. Relative flow distribution of fuel assemblies in reactor core Average temperature of cold pipe section The coolant temperature at the fuel assembly outlet of the reactor core is calculated using the following formula. : 。 4. The method for calculating the average temperature correction of the primary loop heat pipe section of a reactor as described in claim 1, comprising the following steps: Step 3 is to obtain the temperature values ​​of each measuring point in each loop heat pipe section. Establish temperature distribution ranges according to the following rules. : 。 5. The method for calculating the average temperature correction of the primary loop heat pipe section of a reactor as described in claim 1, comprising the following steps: Step 6 is to calculate the temperature offset of each measuring point of the heat pipe section using the following formula: 。 6. The method for calculating the average temperature correction of the heat pipe section in the primary loop of a reactor as described in claim 1, comprising the following steps: Step 7 is to calculate the weighted average temperature of the heat pipe section for each primary loop: 。 7. The method for calculating the average temperature correction of the heat pipe section in the primary loop of a reactor as described in claim 1, comprising the following steps: Step 8 is to obtain the total coolant flow rate in the loop. Use the water property parameter table to look up the enthalpy value of the loop cooling pipe section. and enthalpy of heat pipe section For each primary loop heat, calculate the loop heat power: 。 8. The method for calculating the average temperature correction of the primary loop heat pipe section of a reactor as described in claim 1, comprising the following steps: Step 9 is to add the power of each loop to obtain the total reactor power: 。

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