JPS6132639B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a nuclear reactor simulator using an online computer.

Certain limiting conditions are placed between the thermal output of the reactor and the coolant flow rate. This is to remove the heat generated in the furnace with a sufficient amount of coolant and prevent the fuel assembly from overheating. That is, when operating a nuclear reactor at a certain thermal output, it is necessary to flow a certain amount or more of coolant into the reactor core.

Therefore, when starting up a reactor, restarting it after a planned shutdown, or changing the pattern of control rods, the sequence of operations such as control rod withdrawal and changing the core flow rate must be adjusted so that these limiting conditions are met during the entire operation process. is determined. This operation plan is created in advance by off-line calculation. However, this operation plan is applied with some changes in the following cases.

(1) When the plan does not proceed as scheduled (2) When an unexpected scram (emergency automatic stop) occurs When the operation plan is changed, the new operation plan must comply with the above-mentioned limiting conditions. It is necessary to confirm that it is executable within the scope.

Conventionally, on-site engineers estimate changes in core conditions by referring to past operating results or offline calculation results analyzed for similar cases. However, with this method, it is difficult to accurately estimate the state of the core after operation, that is, changes in core thermal output due to time changes in xenone concentration, changes in nuclear properties due to progress of combustion, etc. Therefore, in order to operate the furnace safely, it is necessary to allow for an excessive amount of thermal margin, resulting in poor operating efficiency.

An object of the present invention is to provide an online nuclear reactor simulator that eliminates the drawbacks of the prior art described above and can predict the state of the reactor core after operation in a short time and with high accuracy.

Hereinafter, a case where the present invention is applied to a boiling water reactor will be explained as an example.

As a basic equation to describe the power distribution in the reactor, Eq.
An approximate neutron diffusion model such as (1) can be used. Equation (1) is a three-dimensional nuclear thermal-hydraulic characteristic analysis program for boiling water reactors called "FLARE".
This is a one-dimensional version of the (FLARE) calculation model (one-dimensional with respect to the core axis direction).

S(K)=k _∞ (K) [S(K-1)・W ^V (K-1)+S(K+1)・W ^V (K+1) +S(K)・W ^V (K)・α ^V (K )+{1−2W ^V (K)−4W ^H (K)+W ^H (K)α ^H (K)}S(K)] (1) Where, K: Node number indicating the axial position S(K ): Power density at node K k _∞ (K): Infinite multiplication factor of neutrons at node K W ^V (K): Vertical direction (axial direction) neutron transport nucleus (neutrons generated at node K are Probability that neutrons will be absorbed by one adjacent node) W ^H (K): Horizontal neutron transport nucleus (Probability that neutrons generated at node K will diffuse horizontally and move outside the node) α ^V (K ): Axial albedo (non-zero values only at nodes at the upper and lower ends of the core) α ^H (K): Horizontal albedo Neutron transport nuclei W ^H (K) and W ^V (K) are functions of void fraction . Furthermore, albedo α ^V (K) and α ^H (K) are constants determined in advance by off-line calculation.

When the in-core power distribution S(K) is given from measured data, the corresponding void distribution can be calculated. From this void distribution, the neutron transport nucleus W ^H
By calculating (K) and W ^V (K) and substituting them into equation (1) together with S(K), we can determine the neutron infinite multiplication factor distribution K _∞ (K) that fits the measured current output distribution. be able to. This is expressed as k ^E _∞ (K).

On the other hand, the infinite neutron multiplication factor can be calculated theoretically from fuel characteristic data such as fuel enrichment, burnup, and void fraction. If we express this as k ^C _∞ (K), we get
k ^C _∞ (K) can be calculated using a calculation formula such as equation (2).

k ^C _∞ (K)={c ₀ +c ₁ U(K)+c ₂ U(K) ² }・{1+δk _Xe +δk _dpp }{1+δk _exp } (2) Here, c ₀ , c ₁ , c ₂ : Coefficient U(K) determined by the control rod insertion state: Moderator density at node _K δk _Xe : Zenon concentration dependent term δk _dpp : Doppler effect dependent term δk _exp : _Burnup dependent _term It can be expressed by a linear equation, a quadratic equation, or the like regarding the control rod insertion ratio r. formula
(3) is an example.

c ₀ = a ₁ + a ₂ r (3) where a ₁ , a ₂ : constant r : control rod insertion ratio k ^c _∞ (K) calculated using equation (2) is the actual measured output mentioned earlier. It has a value different from k ^E _∞ (K) determined from the distribution.
Therefore, in order to resolve the discrepancy between k ^C _∞ (K) based on the calculation model and k ^E _∞ (K) based on the measured value, equation (2) was changed to
Modify as in (4).

k _∞ (K)={c ₀ +δc ₀ (K)+c ₁ U(K)+c ₂ U(K) ² }{1+δk _xe +δk _dpp }{1+δk _exp } (4) Here, δc ₀ (K) is This is a correction term for adapting the calculation model to the actually measured output distribution, and its value is determined by equation (5).

The calculation formula (4) for the infinite neutron multiplication factor determined in this way is a calculation model fixed to the actually measured output distribution. In other words, the simultaneous equations of equation (4) and equation (1) are
Represents a modified computational model describing current core conditions. This model can be used to simulate the reactor core after operation.

Hereinafter, the present invention will be specifically explained with reference to Examples.

FIG. 1 shows the equipment configuration of a nuclear reactor simulator according to the present invention. The measurement signal measured at the measurement input detection point 2 in the reactor 1 is sent to the process input processing device 3
The data is stored in the process input storage unit 4 in the process computer 5 via the process computer 5. Calculation model correction operation section 6
From this process input, the constant input transferred from the magnetic drum 7, and the nuclear thermal hydraulic property calculation model,
Create a new calculation model that fits the current core state. The magnetic drum 7 includes a constant input storage section 8, a nuclear thermal hydraulic characteristic calculation model storage section 9, and an operation plan storage section 10. The calculation model is modified by the method already described. Core state prediction unit 1
1 simulates the core state after executing the operation plan stored on the magnetic drum 7 using this modified calculation model, and displays the results on the cathode ray tube 15.

The start signal for this device is operator console 1.
2 or input from the periodic pulse generator 14 via the selection switch 13. In addition to the start signal, operation plan change instruction data is input from the operator console 12. This is to simulate the core state when the reactor is operated in a manner different from the existing operation plan stored on the magnetic drum 7. If there is no change in the operation plan and it is desired to start the device periodically, a start signal is input from the periodic pulse generator 14.

Next, an example will be shown in which the operation of a nuclear reactor is simulated using the apparatus according to this embodiment. The target furnace is the second one.
This is the boiling water reactor shown in the figure. FIG. 2 is a cross-sectional view of the reactor core, with reference numeral 16 representing a fuel assembly and reference numeral 17 representing a control rod. A guard string 18 is provided within the furnace. A local power range monitor (Local Power Range) is located at the position of this monitoring string 18.
A Traversing Incore Probe (hereinafter referred to as TIP) and a Traversing Incore Probe (hereinafter referred to as TIP) are installed to measure the power distribution within the reactor.
The LPRM is installed at four locations in the axial direction and constantly monitors the local output at each location. TIP is a detector that is inserted from outside the core, and only when necessary.
The continuous axial power distribution at the location of the monitoring string 18 is measured. Reference numeral 19 denotes a virtual monitoring string, which is not equipped with a detector, but is equivalent to the real monitoring string 18 if the reactor core is symmetrical with respect to the axes 20, 20'.

FIG. 3 shows the results of simulating the output increase path of the boiling water reactor. Curve 21 drawn by the broken line in the figure
is the 20% pump speed curve, and point 22 indicates the start point of the output increase due to flow rate control. In addition, a curve 23 drawn by a solid line shows that the reactor output is increased by flow control for approximately 70 hours from a state where the reactor output is approximately 46% of the rated output and the reactor core flow rate is approximately 39% of the rated flow indicated by point 22. The actual measured output increase path when At time point 25, when the core flow rate reaches 100% of the rated flow rate, the reactor power is approximately 76%. Curve 2 drawn with a dashed line
4 is a prediction result when this output increase path is obtained by predictive calculation in the state of point 22 using the means of this embodiment. As is clear from FIG. 3, the output increase path predicted by this example matches the output increase path determined by actual measurements with good accuracy.

FIG. 4 shows the control rod pattern during the flow rate control period described above. The numbers in the figure indicate the state in which each control rod is withdrawn from the core, with a fully inserted state corresponding to 0 and a fully withdrawn state corresponding to 48. The numerical values between these indicate the pull-out distance proportionally. Control rods without numerical values are fully withdrawn.

As described above, according to the present invention, the state of the reactor core after operation can be determined from the current state of the reactor in a short time and with high precision using an online computer.

[Brief explanation of the drawing]

FIG. 1 is an equipment configuration diagram of a nuclear reactor simulator according to the present invention, FIG. 2 is a cross-sectional view of a core of a boiling water reactor,
Fig. 3 is a power-flow curve diagram showing the results of simulating the output increase path of a nuclear reactor using the device of the present invention, and Fig. 4 is a control rod pattern diagram during the output increase period shown in Fig. 3. .

Claims (1)

[Scope of Claims] 1. A first means for generating an electrical signal representing the current state of the nuclear reactor, at least a one-dimensional nuclear thermal hydraulic characteristics calculation model responsive to the electrical signal and representing the power distribution of the reactor. a second means for adjusting at least one parameter; a third means for controlling activation of the second means; and a calculation comprising a parameter matched to the current furnace condition by the second means. An online nuclear reactor simulator comprising a fourth means for generating in real time, based on a model, an electrical signal representing a power distribution after changes in operating conditions that change the power distribution. 2. In the device according to claim 1,
The third means comprises an operator console, a periodic pulse generator, and a changeover switch for selectively applying the output signals of the operator console and the periodic pulse generator to the second means. An online nuclear reactor simulator.