JPH04265899A - Nuclear furnace simulator - Google Patents

Abstract

PURPOSE:To enable a calculation time of a reactor core simulator to be reduced by considering an output of a target unclear power reaching in ALPHAt after a time t to be a known quantity and a change in neutron flux distribution between the time t and (t+ALPHAt) as a function of charge rate of atomic power output. CONSTITUTION:A nuclear furnace 1 is operated according to operation conditions which are set by a nuclear furnace control panel 2, is detected by a neutron detector, and is input to an operation data processing device 4 by an operaiton data input device 3. The device 4 allows most recent neutron flux data to be input data recording device 6. Then, with a neutron flux phi(r,t) at a time t of a reactor core combustion calculation, a nuclear constant, and a target nuclear furnace output P at a time (t+ALPHAt) as known quantities, a transition change of the distriburion phi(r,t) within a combustion time step can be obatined by phi(r,t+ALPHAt)=phi(r,T)X(1+mXALPHAt) using a relative change rate m of the nuclaer furnace output within the combustion time step between the time t and (t+ALPHAt).

Description

[Detailed description of the invention]

0001

[Industrial Application Field] The present invention relates to a nuclear reactor simulator for managing the operation of a nuclear reactor, and is particularly useful for improving control rod planning for daily load operation of a nuclear reactor and restart operation after an emergency shutdown. This invention relates to a preferred nuclear reactor simulator.

0002

[Prior Art] Generally, the startup operation of a BWR or PWR is
This is done based on a power change plan (plan regarding time changes in reactor power and control rod pattern) prepared in advance by core management engineers. On the other hand, if you want to change the reactor output operation plan each time depending on the power demand situation, plant site engineers can use an output change planning support system that creates a load following plan to plan the output change. Create. However, in a power change plan that involves control rod manipulation, it is necessary to use a reactor simulator or the like to evaluate core operating characteristics such as thermal limit values during the power change. When creating a restart plan after an emergency reactor shutdown, it is necessary to quickly assess how fission products such as xenon and samarium accumulate in the reactor core at the time of restart, and select an appropriate control rod pattern. There must be. In addition, considering the fact that the burnup of nuclear reactor fuel is increasing and fuels with different neutron spectra are mixed in the reactor, there is a high need for a nuclear reactor simulator that can create a restart plan after an emergency reactor shutdown. . Therefore, it is an important issue to improve both the calculation accuracy and calculation time of nuclear reactor simulators.

By the way, the temporal change in the atomic number density of a fission product (for example, xenon) with a large thermal neutron absorption cross section, which is produced in nuclear fuel placed in a nuclear reactor, is determined, and according to the change, As an example of a method of operating a nuclear reactor that increases the output using an output control means,
There is a publication No. 2097. In this conventional example, the xenon atomic number density Nxe(t+Δt
) is the neutron flux φ(t
) and the xenon atomic number density Nxe(t) after t time were used. Nxe(t+Δt)=f(φ(t), Nx
e(t))
...(1)

0004

[Problem to be solved by the invention] In the conventional example, transient changes in the three-dimensional distribution of xenon atomic number density in the core are not considered, so thermally important values such as power peaking satisfy the design constraint value. There was a problem in that it was not possible to accurately understand what was going on, and it was not possible to create highly accurate control rod plans.

On the other hand, in conventional nuclear reactor simulators based on three-dimensional models that are calculated off-line, t~t+Δt
There is a method in which the neutron flux distribution after time t + Δt is updated sequentially to reflect the reactivity effect due to fission products that absorb thermal neutrons, which accumulates during the period, and to converge including the thermal-hydraulic conditions. It is taken. Therefore, since detailed prediction calculations must be performed off-line, creating a restart operation plan requires a lot of time and time, and there is a problem in that the accuracy is not good.

As mentioned above, it will be important in the future to improve both calculation accuracy and calculation time for nuclear reactor simulators based on three-dimensional multi-group models of reactor cores loaded with high-burnup fuel, which will become mainstream in the future. It is expected that the frequency will increase.
This is an important issue in order to provide a core management program and a core simulator that can support the creation of daily load operation and restart operation plans.

The purpose of the present invention is to shorten the calculation time and improve the accuracy of a three-dimensional multi-group core management program and a core simulator necessary for creating plans for daily load following operation and restart operation after emergency shutdown. It's about trying. .

[0008]

[Means for Solving the Problems] The present invention is characterized in that the atomic number density of fission products in the fuel and the power distribution within the core are calculated according to the method shown below. That is,
(i) From a certain time t, assuming that the target reactor power to be reached after a time step Δt is a known quantity, the reactor power P(t) at time t and the target reactor power P(t+
Δt) is the input value.

(ii) The change in the neutron flux distribution between time t and t+Δt is the rate of change m in the reactor power P, m=[(P(t+Δt)−P(t))
/P(t)]/Δt...(2)
Expressed as a function related to

φ(r,t+Δt)=g(φ(r,t
), Δt, m) …(
3) When changing the reactor power in a ramp-like manner, considering that the neutron flux also changes at the same rate of change as the reactor power, equation (3) becomes φ(r, t + Δt) = φ(r, t )×
(1+m×Δt)…(4)
It is written as

(iii) The change in the atomic number density of the fission product between time t and t+Δt is expressed by the differential equation of the atomic number density, dN(r,t)/dt=h(
N(r, t), φ(r, t)) …(
5) Calculate using the analytical solution obtained when the neutron flux in is the function given in (ii) above.

(iv) Calculate the neutron flux and power distribution at time t+Δt by considering the atomic number density distribution of fission products at time t+Δt.

[0013]

[Operation] For example, when increasing the reactor output between a certain time t and t+Δt, the magnitude of the neutron flux also changes in proportion to the rate of increase in the reactor output. When the size of the neutron flux reaches the order of 1013,1014 (n/cm2・sec), fission products with a large thermal neutron absorption cross section, such as Xe-135, accumulate in the reactor core. This will affect the reactivity. At this time, Xe-
The time change in the atomic number density of 135 is approximately proportional to the time integral value of the thermal neutron flux, and the influence of short-term transient changes in the thermal neutron flux is small. Therefore, using the neutron flux distribution based on the neutron flux distribution at time t, the spatial distribution of the atomic number density of xenon formed at the final time t+Δt is determined. In the present invention, the rate of change m in the reactor output between time t and t+Δt is used as an index indicating the temporal change in neutron flux. That is, from a certain time t,
By considering the change in neutron flux distribution between time t and t + Δt as a function regarding the rate of change m of the reactor power, assuming that the target reactor power to be reached after time step Δt is a known quantity, nuclear fission generation between time t and t + Δt is achieved. Changes in the atomic number density of a substance over time can be determined by analytically solving its differential equation. Since it can be calculated based on the analytical solution of the differential equation, calculation errors in the process of calculating the atomic number density of fission products can be reduced. Further, based on the atomic number density distribution of fission products at time t+Δt, which is directly obtained by one calculation, the neutron flux distribution of the target reactor output can be obtained by one direct calculation. Therefore, the calculation time of the nuclear reactor simulator can be reduced.

[0014]

[Example] Examples will be explained below.

FIG. 1 shows a calculation algorithm for determining the atomic number density distribution and power distribution of fission products at time t+Δt, which is Δt after a certain time t. First, the target reactor output at time t+Δt is input. Thereafter, at time t+Δt, the atomic number densities of fission products with a large thermal neutron absorption cross section and nuclides that produce fission products by decay are calculated. In this example, the entire core system including the reflector section is treated with three-group neutron energy and three-dimensional diffusion calculations, and Xe-135, Sm-149, and their production are treated as fission products with large thermal neutron absorption cross sections. As a fission product related to I-13
5, deals with transient changes in Pm-149. Also, time t+
The neutron flux φi (r, t + Δt) after Δt is calculated as φi (r, t + Δt) = φi (r, t) × (1
+m×Δt)…(6)
(i=1, 2, 3) Here, i=1 (high speed group), i=2 (medium speed group), i=3
(thermal group).

[0016] I-135, Xe-135, Pm-149
, the atomic number density of Sm-149 is expressed by their differential equation,

[0017]

[Math 1]

It can be obtained directly by one calculation from the analytical solution obtained from ##EQU1## At time t+Δt, Xe-135 and Sm-149, which are nuclides with large thermal neutron absorption cross sections,
Solve the neutron diffusion equation numerically by reflecting the macroscopic absorption cross section of .

FIG. 2 shows a non-boiling light water-cooled research atom with a neutron flux density of 1014 (n/cm2·sec) class using a core combustion calculation program based on a three-dimensional three-group diffusion model to which the method of the present invention is applied. This is an example of performance analysis of a 100-hour rated power operation of a reactor, and shows the change in reactivity over time based on a clean core. The measured values were obtained using control rod calibration curves obtained through experiments, and show very good agreement with the calculated values obtained in this example. That is, using the core combustion calculation program based on the present invention,
It is possible to accurately predict the critical control rod pattern of a nuclear reactor. Further, as shown in FIG. 2, there is an operation method in which a nuclear reactor is started up according to a control rod plan planned by a core burnup calculation program based on the method of the present invention. FIG. 2 is an example showing a change in reactivity over time compared to actual measurement based on predictive analysis. In the present invention, the time required for iterative calculation of the atomic number density distribution of fission products can be shortened, and the time required for iterative calculation of the atomic number density distribution of fission products can be reduced by approximately
0% can be shortened. In the above embodiment, the rate of change m of the reactor power may be determined using the core average neutron flux that is proportional to the reactor power.

The method of the invention described in the above embodiments can also be applied to online calculations. As an example,
FIG. 3 shows the basic configuration of an online nuclear reactor simulator to which the method of the present invention is applied. In FIG. 3, the reactor 1 is
The reactor is operated under operating conditions set by the reactor control panel 2. It is detected by a neutron detector (not shown), and is input to the operation data processing device 4 by the operation data input device 3. The operation data processing device 4 stores the latest neutron flux data in the data storage device 6, and based on this neutron flux data, from the time t when this data is adopted, to the time t.
Assuming that the neutron flux level during +Δt is constant, Xe-13
5, Sm-149, etc., is estimated by the operating state estimating device 5, and at the same time, the neutron flux distribution and power distribution at time t+Δt are estimated using the obtained atomic number density. The estimation result is stored in the data storage device 6, and the obtained reactor operating state is determined by the limit condition determination device 7.
It is compared with limiting conditions determined from the operating state, and the results are displayed on the display device 8.

[0021]

[Effects of the Invention] By using the core combustion management program or core simulator to which the present invention is applied, calculation time is shortened, and control rod plans for operations involving output changes such as reactor restart operations are created with high accuracy. can do.

[Brief explanation of the drawing]

FIG. 1 is an explanatory diagram of a calculation algorithm of the present invention.

FIG. 2 is an explanatory diagram of an example of analysis by a core combustion management program to which the calculation algorithm of the present invention is applied.

FIG. 3 is a block diagram of a nuclear reactor simulator to which the calculation method of the present invention is applied.

[Explanation of symbols]

1... Nuclear reactor, 2... Nuclear reactor control panel, 3... Operating data input device, 4... Operating data processing device, 5... Operating state estimation device,
6...Data storage device, 7...Limiting condition determination device, 8...Display device.

Claims (9)

[Claims] Claim 1: In a nuclear reactor simulator that predicts the core operating characteristics of a nuclear reactor in order to create a control rod plan for an operation that involves a change in reactor output such as a restart operation of the reactor, a time t of core burn-up calculation is used. Assuming that the neutron flux φ(r, t), the nuclear constant, and the target reactor power P at time t+Δt are known quantities, and using the relative change rate m of the reactor power during the combustion time step between time t and t+Δt, The transient change in the neutron flux distribution φ(r, t) during the combustion time step is expressed as φ(r,
t + Δt) = φ (r, t) × (1 + m × Δt), where the neutron flux φ (r , t) from the analytical solution obtained when the function form is used, the atomic number density of the nuclide is calculated, and the neutron flux distribution and power distribution at time t + Δt are determined by considering the calculated atomic number density distribution. A nuclear reactor simulator featuring steps. 2. In claim 1, a relative change rate m of the reactor power during the combustion time step between time t and t+Δt.
A nuclear reactor simulator that calculates from the core average neutron flux proportional to the reactor power. Claim 3: In claim 1, the time t of core combustion calculation.
A nuclear reactor simulator that calculates the atomic number density of the nuclide from an analytical solution of a differential equation of the atomic number density obtained by representing a transient change in the neutron flux distribution φ(r, t) between ~t+Δt as a constant neutron flux distribution. 4. A nuclear reactor simulator that calculates transient changes in the atomic number density distribution of a nuclide with a very large thermal neutron absorption cross section in a nuclear reactor outside of a neutron flux distribution iterative calculation process. 5. In claim 1, 2 or 3, using the output instead of the neutron flux in the differential equation of the atomic number density,
A nuclear reactor simulator that calculates the atomic number density of the nuclide. 6. Calculating the atomic number density of the nuclide from an analytical solution obtained under the constant neutron flux distribution conditions according to claim 3, based on a neutron density signal obtained from a neutron detector installed in a nuclear reactor. An online reactor simulator that predicts changes in core operating characteristics. 7. In the nuclear reactor simulator according to claim 1, 2, 3, 4, 5, or 6, in applying the analytical solution for determining the atomic number density of a nuclide with a very large thermal neutron absorption cross section, A nuclear reactor simulator that is capable of selecting an analytical solution by setting determination conditions such as m≠0 and m=0 regarding the relative rate of change m of the reactor output described above. 8. The nuclear reactor simulator according to claim 1, 2, 3, 4, 5, 6, or 7, wherein numerical analysis is applied instead of an analytical solution to calculate the atomic number density. 9. A nuclear reactor operating method according to claim 1, 2, 3, 4, 5, 6, 7, or 8, which operates a nuclear reactor based on a prepared daily load operation plan or restart operation plan.