JP2001074875A - Method for obtaining temperature difference between inlet/outlet of reactor vessel of pressurized water reactor, and method for evaluating performance of pressurized water reactor plant - Google Patents

Abstract

PROBLEM TO BE SOLVED: To improve reliability when detecting the temperature difference between the inlet/outlet of the reactor vessel of a pressurized water reactor. SOLUTION: The correlation between a measurement reactor core outlet temperature distribution obtained by a detector 57 of temperature in the reactor being arranged at the outlet of a reactor core 47 in a reactor vessel 31 and a calculation reactor core outlet temperature distribution being calculated based on the reactor core output distribution obtained by a detector 55 of neutrons in the reactor is obtained, thus calculating the temperature of the outlet of the reactor core of a cooling material. Assuming that there is no heat transfer from the outlet of the reactor core to that of the reactor vessel, the calculated temperature is set to the temperature of the outlet of the reactor vessel.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

[0001] The present invention relates to a method for evaluating the performance of a nuclear reactor, in particular, a pressurized water reactor.

[0002]

2. Description of the Related Art In a pressurized water reactor, a difference between a reactor vessel outlet temperature of a reactor coolant and a reactor vessel inlet temperature is used as an important parameter for operation control and performance evaluation. In summary, in FIG. 4, which shows a typical system diagram of a pressurized water nuclear power plant, a coolant heated in a core 3 in a reactor vessel 1 passes through a high-temperature side pipe or hot leg 5 and a steam generator 7. Then, it exchanges heat with the feed water as described later to convert it into steam, which itself is cooled down and sucked into the coolant pump 11 via the intermediate pipe or crossover leg 9. The coolant pump 11 returns the cooled coolant to the reactor vessel 1 through the low-temperature side pipe, that is, the cold leg 13, and the coolant is heated again in the core 3. The system in which the coolant circulates in this manner is also called a primary cooling system. The pressure of the primary cooling system is controlled by a pressurizer 15 to a predetermined range. As described above, the steam generated by being heated by the reactor coolant in the steam generator 7 is sent to the steam turbine 23 through the main steam pipe 21,
Here, the generator is driven to generate electricity. Exhaust steam of the steam turbine 23 is cooled and condensed in a condenser 25 to be condensed water, and is again supplied as water to the steam generator 7 by a secondary coolant pump to a water supply pump 27. During operation, the above-mentioned circulation system, that is, the secondary cooling system is circulated.

The amount of heat supplied to the feed water from the reactor coolant in the steam generator 7, ie, the reactor heat output, the pressure of the steam in the steam generator 7, ie, the main steam pressure, and the power generation connected to the steam turbine 23 The electrical output of the machine is a major indicator in operation control or performance evaluation. The thermal output of the reactor is determined by the temperature detector 17 provided in the hot leg 5 and the reactor vessel outlet temperature T _HOT of the coolant and the cold leg 13.
Is obtained as the product of the difference between the coolant and the reactor vessel inlet temperature T _IN by the temperature detector 19 provided at the reactor vessel, that is, the reactor vessel inlet / outlet temperature difference ΔT and the flow rate of the reactor coolant. The temperature difference ΔT between the entrance and exit of the reactor vessel is determined by the temperature detector 17 and the temperature detector 1.
In the constant flow rate operation, which is obtained as the output of the subtractor 18 connected to 9 and does not substantially change the coolant flow rate, the reactor vessel inlet / outlet temperature difference ΔT is used as an important parameter in plant performance evaluation or operation control. .

In order to monitor the power distribution in the reactor core 3 in the reactor vessel 1 during operation, a plurality of reactor temperature detectors or thermocouples are provided at the outlet of the reactor core 3 and the reactor core is provided. The in-furnace neutron detector is arranged in a specific fuel assembly constituting 3. FIG. 5 shows the arrangement of the in-furnace temperature detector 26 and the in-furnace neutron detector 28 in one quadrant on the upper surface of the core 3 (the in-furnace temperature detector 26 and the in-furnace neutron detector with respect to the central axis of the core 3). The arrangement of 28 is usually rotationally symmetric.) In the figure, reference numeral 29 denotes a fuel assembly, and as can be seen from the figure, the fuel assembly 29 is provided with only the in-furnace temperature detector 26, provided with only the in-furnace neutron detector 28, Some are provided with both the internal temperature detector 26 and the in-furnace neutron detector 28, and some are not. The furnace temperature detector 26
Is not provided directly on the fuel assembly 29, but is provided on the upper core plate side facing the upper end thereof.

[0005]

The above-mentioned coolant outlet temperature T _HOT of the coolant is obtained as an average value of several temperature detectors 17 arranged in the large-diameter hot leg 5. However, there is a problem in reliability because it is affected by various factors. Therefore, an object of the present invention is to provide a method for obtaining a highly reliable reactor vessel inlet / outlet temperature difference ΔT which is not affected by other factors.

[0006]

According to the method of the present invention, the temperature difference between the inlet and outlet of a reactor vessel of a pressurized water reactor is determined by the core outlet temperature distribution obtained by an in-core temperature detector. From the core power distribution by the sub-channel thermal analysis code using the core power distribution obtained in the reactor neutron detector, determine the respective temperature rise width at each installation position of the furnace temperature detector, using those values A correlation equation between the core outlet temperature distribution by the in-core temperature detector and the core power distribution by the sub-channel thermal analysis code is obtained, and the average core temperature is calculated from the average core outlet temperature by the sub-channel thermal analysis code using the correlation equation. The reactor vessel outlet temperature is determined in consideration of the core bypass flow rate, and the reactor vessel outlet temperature is calculated from the reactor vessel outlet temperature. Obtained by subtracting the reactor vessel inlet temperature. Further, according to the present invention, preferably, the correlation equation used in carrying out the above method determines the difference between the measurement result of the in-furnace temperature detector and the analysis result by the sub-channel thermal analysis code, and calculates the difference. Evaluate using a statistical method, determine the application range, such that the difference between the measurement result of the furnace temperature detector and the analysis result by the sub-channel thermal analysis code is within the application range, It is obtained using the measurement result of the temperature detector and the analysis result by the sub-channel thermal analysis code. Then, the performance of the pressurized water reactor is evaluated using the reactor vessel inlet / outlet temperature difference obtained as described above.

[0007]

Embodiments of the present invention will be described below with reference to the accompanying drawings. FIG. 1 conceptually shows a structure in a reactor vessel of a pressurized water reactor 30 for which a reactor vessel inlet / outlet temperature difference ΔT is to be obtained by the method of the present invention. The reactor vessel 31 has a reactor vessel body 33 and a vessel lid 35, and the reactor vessel body 33 has a coolant inlet nozzle 37 and an outlet nozzle 39 integrally formed. A core tank 41 is suspended from the upper opening of the reactor vessel body 33 and defines an annular descending flow path 43 communicating with an inlet nozzle 37. I support it. Further, an upper core plate 51 integrally connected to an upper core support plate 49 supported by the upper opening is located on the outlet side of the core 47. The in-furnace neutron detectors 55 can be inserted into and removed from a plurality of thimble guide tubes 53 only showing two tubes, respectively, while a plurality of temperature detectors 57 showing only one tube are arranged in the upper core plate 51. This extends to the temperature detector port 59 and extends out of the reactor vessel therefrom. The planar arrangement of the plurality of in-furnace neutron detectors 55 and temperature detectors 57 in the core is the same as that in FIG. 5 described above.

[0008] The flow of the coolant in the pressurized water reactor 30 having the above-described structure will be described. Flows into. Then, the coolant changes the flow direction upward in the lower plenum 61,
Dispersed inside the core 47 and flows upward. More specifically, the fuel rods 45 flow inside and outside of the fuel assembly 45 in contact with the outer surfaces of the fuel rods and flow upward in parallel with the fuel rods, and take up the heat of nuclear reaction to increase the temperature. Although not shown, a forming plate region is provided on the outer peripheral portion of the core 47, and a small amount of coolant flows there and rises. The coolant that has flowed upward and downward in and out of the core 47 merges and mixes in the upper plenum 63 and flows out to the hot leg through the outlet nozzle 39. The temperature detector 57 detects the coolant temperature at the outlet of the corresponding fuel assembly 45. The in-core neutron detector 55 detects a neutron flux in the fuel assembly 45 in the core 47 into which it is inserted. Temperature detector 5
7 is substantially proportional to the power distribution in the horizontal direction of the core 47.

FIG. 2 shows the basic principle of the method of the present invention. Finding a correlation such that the difference between the inlet / outlet temperature difference (ΔT) 65 of the fuel assembly 45 obtained from the core power distribution and the inlet / outlet temperature difference (ΔT) 67 obtained from the temperature measured by the temperature detector 57 is minimized. Thus, a more reliable core entrance / exit temperature difference (ΔT) can be obtained. The procedure for obtaining the reactor vessel inlet / outlet temperature difference (ΔT) will be described below. 1. The core outlet temperature (T _HOT ) at each installation position of the temperature detector 57 is determined from the core power distribution using the sub-channel thermal analysis code. 2. The measured value of the temperature detector 57 (T _{HOT T / C} ) and the core outlet temperature calculated value (T
By subtracting the reactor inlet temperature (T _IN ) from _{HOT CAL} ), the fuel assembly inlet _/ outlet temperature difference (ΔT _{T / C} ) based on the measured value at the installation position of each temperature detector 57 and the subchannel thermal analysis code base The fuel assembly inlet / outlet temperature difference (ΔT _CAL ) is determined. 3. Fuel assembly inlet and outlet temperature difference ([Delta] T T _{/ C)} and the fuel assembly inlet and outlet temperature difference ([Delta] T _{C AL)} determines the correlation equation between them by a linear expression shown because of the one-to-one relationship (1). y = ax + b (1) where, y: measurement result of fuel assembly inlet _/ outlet temperature difference (ΔT _{T / C} ) x: analysis result of fuel assembly inlet _/ outlet temperature difference (ΔT _CAL ) a: slope b: intercept 4 . The reactor inlet / outlet temperature difference (ΔT _{AVG CAL} ) based on the sub-channel thermal analysis code is obtained by subtracting the reactor inlet temperature (T _IN ) from the average core outlet temperature (T _{H AVG CAL} ) obtained by the sub-channel thermal analysis code. Ask for 73. 5. The core inlet / outlet temperature difference (ΔT _{AVG CA L} ) obtained in the fourth term is substituted into the correlation equation obtained in the above three terms, and as shown in FIG. 3, the core inlet / outlet temperature difference (ΔT _{cor e cal} )
Find 71. 6. By adding the reactor vessel inlet port temperature T _IN to the core inlet / outlet _port temperature difference (ΔT _{core cal} ) 71, the core outlet port temperature (T
_{H0T PC} ) is obtained. 7. The aforementioned core outlet temperature (T _{H0T PC),} obtaining the consideration of core bypass flow rate reactor vessel outlet temperature (T _{H0T P).} 8. The reactor vessel inlet / outlet temperature difference (ΔT _P ) is obtained by subtracting the reactor vessel inlet temperature T _IN from the aforementioned reactor vessel outlet temperature (T _{H0T P} ).

Further, in the present invention, it is possible to further improve the prediction accuracy of the reactor vessel inlet / outlet temperature difference by performing the following procedure. 9. As described above, the in-furnace temperature detector 57 and the in-furnace neutron detector 55 are not provided on all of the fuel assemblies 45 of the core 47, and are arbitrarily arranged so as to cover substantially the entire core 47. . Therefore, the fuel assembly 4 provided with both the furnace temperature detector 57 and the furnace neutron detector 55 is provided.
If the correlation of the core outlet temperature is obtained only with the reference numeral 5, it is possible to reduce the error due to the difference between the measurement positions of the in-furnace temperature detector 57 and the in-furnace neutron detector 55. 10. The core outlet temperature calculated value (T) using the measured value (T _{HOT T / C} ) of the temperature detector 57 and the sub-channel thermal analysis code
_In order to extract the fuel assembly 45 having a good correlation with the _{HOT CAL} ), the measured value (T _{HOT T / C} ) of the temperature detector 57 and the core outlet temperature calculated value (T
_{HOT CAL} ), the absolute value (| ΔT _S |) of the difference (ΔT _S |) is obtained, and the standard deviation (σ) of the absolute value (| ΔT _S |) of the difference is obtained. The evaluation use range (| ΔT _S | _R ) is the average value (| Δ
T _s | _AVG ) and the standard deviation (σ) described above. The measured value (T _{HOT T / C} ) of the temperature detector 57 of the fuel assembly 45 within the evaluation use range (| ΔT _S | _R ) and the calculated value of the core outlet temperature by the sub-channel thermal analysis code
(T _{H OT CAL)} only, the reactor vessel inlet and outlet temperature difference ([Delta] T
_P ) is determined with higher accuracy. Note that a probability error (0.6745σ) or a limit range using other statistical methods may be used instead of the standard deviation (σ). Further, in the above-described method for obtaining the reactor vessel inlet / outlet temperature difference (ΔT _P ), the temperature is used as a parameter. However, the accuracy is further improved by calculating the enthalpy as a parameter instead.

In the above-mentioned correlation equation, the reactor vessel inlet temperature T _IN
Then, the core outlet temperature (T _HOT ) at each installation position of the temperature detector 57 can be obtained by the following equation. y _t = ax + b + T _IN (2) where, y _t : core outlet temperature (T _HOT ) x: analysis result of fuel assembly inlet / outlet temperature difference (ΔT _CAL ) a: slope b: intercept T _IN : atom Reactor Vessel Inlet Temperature As described above, the core vessel average inlet / outlet temperature difference (ΔT
By substituting ( _{CORE CAL} ) into x in the equation (2), the core average core exit temperature (T _{H0T PC} ) can be obtained.

By transforming the above-mentioned correlation equation (1) into the following equation, even when the measurement by the in-furnace neutron detector 55 is not performed, the reactor The internal power distribution can be determined. z = (y−b) / a × α (3) where, z: in-furnace power distribution obtained from the measurement result of the in-furnace temperature detector 57, y: fuel assembly inlet / outlet temperature difference obtained from equation (1) (ΔT
_{T / C} ) a: Correlation equation constant b: Correlation equation constant α: Conversion coefficient

In the evaluation of the plant performance of the pressurized water reactor described above, the heat output, which is the amount of heat exchanged in the steam generator, is a main evaluation item. The reactor vessel inlet / outlet temperature difference (ΔT _P ) is used instead of the reactor vessel inlet / outlet temperature difference ΔT, and a more reliable evaluation can be made. Further, by obtaining the reactor vessel entrance / exit temperature difference (ΔT _P ), the reliability of the reactor vessel entrance / exit temperature difference ΔT used conventionally can be guaranteed.

[0014]

As described above, according to the present invention,
Since the average coolant temperature at the reactor core outlet is obtained from the core outlet temperature of the coolant by the reactor temperature detector provided in the reactor vessel and the reactor power distribution by the reactor neutron detector,
A more accurate and reliable reactor vessel inlet / outlet temperature difference can be obtained. Further, according to the present invention, it is possible to statistically process the core outlet temperature of the coolant by the in-furnace temperature detector and the calculated core outlet temperature obtained from the in-core power distribution by the in-furnace neutron detector. It is possible to determine the temperature difference between the reactor vessel inlet and outlet. Further, according to the present invention, since the accurate reactor vessel inlet / outlet temperature difference is used as described above, accurate performance evaluation can be performed.

[Brief description of the drawings]

FIG. 1 is a conceptual diagram of the interior of a reactor vessel of a pressurized water reactor in which a temperature difference between a reactor vessel entrance and an exit is obtained by a method of the present invention.

FIG. 2 is a basic principle diagram of the present invention.

FIG. 3 is an operation explanatory view of the present invention.

FIG. 4 is a typical system diagram of a conventional pressurized water nuclear power plant.

FIG. 5 is a conceptual diagram showing the arrangement of a reactor temperature detector and a reactor neutron detector in the pressurized water nuclear power plant of FIG.

[Explanation of symbols]

 REFERENCE SIGNS LIST 30 pressurized water reactor 31 reactor vessel 33 reactor vessel body 35 vessel lid 37 inlet nozzle 39 outlet nozzle 41 core vessel 43 annular descending flow path 45 fuel assembly 47 core 49 upper core support plate 51 upper core plate 53 thimble guide tube 55 In-furnace neutron detector 57 In-furnace temperature detector 61 Lower plenum 63 Upper plenum

 ────────────────────────────────────────────────── ─── Continuing on the front page (72) Inventor Akihiko Tominaga 708-1 Otauemachi, Takamatsu City, Kagawa Prefecture Apartment of Shiden Ota (72) Inventor Toshiaki Niihara 1-1-1, Wadazakicho, Hyogo-ku, Kobe City, Hyogo Mitsubishi Heavy Industry Co., Ltd. Kobe Shipyard F-term (reference) 2G075 AA05 BA03 CA08 CA40 DA01 DA03 FA11 FA19 FB07 FB15 FC06 GA21

Claims (3)

[Claims] 1. A core outlet temperature distribution obtained by a subchannel thermal analysis code using a core outlet temperature distribution obtained by an in-core temperature detector and a core power distribution obtained by an in-core neutron detector. The respective temperature rise widths at the respective installation positions are obtained, and the correlation equation between the core outlet temperature distribution by the in-furnace temperature detector and the core power distribution by the sub-channel thermal analysis code is obtained by using those values. The average temperature rise of the core is determined from the average temperature of the core outlet by the sub-channel thermal analysis code, and the reactor vessel inlet temperature is added to the average temperature rise by adding the reactor vessel inlet temperature and the core bypass flow rate. And calculating the reactor vessel inlet / outlet temperature difference of the pressurized water reactor by subtracting the reactor vessel inlet temperature from the reactor vessel outlet temperature. 2. The difference between the measurement result of the in-furnace temperature detector and the analysis result by the sub-channel thermal analysis code is obtained, and the difference is evaluated using a statistical method to obtain an applicable range.
The difference between the measurement result of the furnace temperature detector and the analysis result by the sub-channel thermal analysis code is within the applicable range, and the measurement result of the furnace temperature detector and the analysis result by the sub-channel thermal analysis code are 2. The method for determining a temperature difference between a reactor vessel inlet and an outlet of a pressurized water reactor according to claim 1, wherein the correlation equation is determined using the correlation equation. 3. A core outlet temperature distribution obtained from a core outlet temperature distribution obtained by an in-core temperature detector and a core power distribution obtained by a subchannel thermal analysis code using a core power distribution obtained by an in-core neutron detector. The respective temperature rise widths at the respective installation positions are obtained, and the correlation equation between the core outlet temperature distribution by the in-furnace temperature detector and the core power distribution by the sub-channel thermal analysis code is obtained by using those values. The average temperature rise of the core is determined from the average temperature of the core outlet by the sub-channel thermal analysis code, and the reactor vessel inlet temperature is added to the average temperature rise by adding the reactor vessel inlet temperature and the core bypass flow rate. Is obtained, the reactor vessel inlet / outlet temperature is subtracted from the reactor vessel outlet temperature to obtain a reactor inlet / outlet temperature difference, and the reactor inlet / outlet temperature difference is used as a performance evaluation index. Performance evaluation method for pressurized water reactor plant.