JPH05107388A - Method for measuring neutron effective multiplication constant during storage of radiation fuel - Google Patents

Abstract

PURPOSE:To readily calculate a neutron effective multiplication constant at a radiation fuel storage portion by substituting known standard fuel assemblies for some fuel assemblies having high reaction effects, and using two calibration curves calculated independently according to the respective neutron flux values of the fuel assemblies. CONSTITUTION:UO2 fuel rods of light water reactor type are arranged in a square lattice and 7X7 fuel assemblies are so arranged that eight fuel assemblies 1b are aligned around a center fuel assembly 1a via two columns of water gaps 2. A neutron monitor 3 is disposed at the center of the lower side of the total of the nine fuel assemblies to measure neutron flux. The assembly 1a is set at a position at which reaction effects are high. The effective fissile concentration of the assemblies 1b is determined using a calibration curve according to the fissile concentration and the neutron flux value of a standard fuel assembly (assembly 1a) of known composition. Then a neutron effective multiplication constant at a fuel storage portion is calculated using the next calibration curve according to the effective fissile concentration of the assemblies 1b and the fissile concentration of the assembly 1a indicated by a parameter.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an irradiation fuel assembly which is loaded into a nuclear reactor and is irradiated with neutrons, and is temporarily taken out of the nuclear reactor during use or becomes a spent fuel. The present invention relates to a method for measuring the effective neutron multiplication factor in a fuel storage area containing a body.

[0002]

2. Description of the Related Art Fuel used in a nuclear reactor is taken out of the core in units of fuel assemblies and temporarily or long-term stored in a fuel storage rack installed in a water pool. Alternatively, after being stored in a fuel storage rack for a certain period of time to attenuate radioactivity having a short half-life, it is stored in a spent fuel transportation container (cask) and sent to a reprocessing facility. The spent fuel sent to the reprocessing facility is temporarily stored in a submerged fuel storage rack.

In countries where the reprocessing of spent fuel is delayed or is not reprocessed, spent fuel is stored in a spent fuel transportation and storage container which also serves as a transportation container and a storage container. Become.

[0004]

The fuel of a nuclear reactor is allowed to be critical in the core of a nuclear reactor having a control function, but if the fuel other than the core is absolutely subcritical when the fuel is stored. I won't. Therefore, conventionally, the fuel storage device must be designed so that the subcriticality can be secured with a sufficient margin for any fuel, from new fuel to spent fuel, even when the neutron multiplication characteristic is the highest. I have to. This is apt to be uneconomical because in an actual state, there is an excessive margin, and the amount of fuel that can be stored in a certain volume range is small. In recent years, due to the delay in reprocessing or the increase in the amount of spent fuel stored, the need to store more fuel in a certain space and reduce the waste of the storage space as much as possible has attracted attention. For this reason, when the fuel is packed densely, the subcriticality margin when the fuel is stored will generally decrease, so we want to ensure criticality safety by confirming the actual subcriticality when storing the fuel. However, this confirmation method has not been established in the past.

An object of the present invention is to store a part of the irradiated fuel assemblies and a standard fuel of a known composition for the irradiated fuel assemblies in the storage portion when a plurality of irradiation fuel assemblies are stored in a subcritical manner. An object of the present invention is to provide a method for measuring the effective neutron multiplication factor during storage of irradiated fuel, which can be replaced with an assembly and easily obtain the effective neutron multiplication factor in the irradiated fuel storage portion from each neutron flux value.

[0006]

A part of the irradiation fuel assembly containing a plurality of irradiation fuel assemblies in a subcritical irradiation fuel storage area where the reactivity effect is high, that is, in a location where the neutron importance is high. Replace the standard fuel assembly of known composition, measure the neutron flux in each, the periphery of the standard fuel assembly using the calibration curve separately obtained from the composition of the standard fuel assembly and the neutron flux value. The effective neutron multiplication factor of the fuel storage unit which stores the irradiation fuel assembly is known by using the calibration curve separately obtained from this composition and the composition of the standard fuel assembly.

[0007]

In the fuel containing section containing a plurality of irradiated fuel assemblies, a part of the irradiated fuel assemblies is replaced with a standard fuel assembly having a known composition, and the neutron flux in each of the fuel containing sections is measured. Then, the effective composition around the standard fuel assembly is obtained from the calibration curve separately obtained from the neutron flux value and the composition of the standard fuel assembly, and the calibration separately obtained from this composition and the composition of the known standard fuel assembly. The curve is used to obtain the neutron effective amplification factor of the fuel storage unit that stores a plurality of irradiated fuel assemblies, thereby ensuring subcriticality.

[0008]

DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described with reference to the drawings. The present invention generally has a high neutron importance in the central portion of the fuel storage portion, and in such a place, a relatively small amount of substance (here, the neutron effective multiplication factor in the fuel storage portion (hereinafter simply referred to as an effective enhancement by replacing the fuel assembly) is used. It is possible to effectively change the magnification (also referred to as a magnification, which is also represented by keff) or the reactivity (used synonymously with the subcriticality ρ = keff-1 here). `` In a storage unit with a small keff value, the change in neutron flux is small even if the neutron multiplication characteristic of the fuel assembly to be replaced is large, but even if the fuel assembly is replaced with the same fuel assembly, the ke
The neutron flux changes significantly if the ff value is large. " If the neutron flux level does not change after the change in neutron emission rate is corrected even if the standard fuel assembly is inserted in place of the extracted irradiated fuel assembly, the neutron absorption characteristics and neutron enhancement of the extracted irradiated fuel assembly The double characteristic is equivalent to the effectively replaced standard fuel assembly. FIG. 1 is a layout view of a fuel assembly showing an embodiment of the present invention. A light water reactor type UO ₂ fuel rod (pellet diameter 1.0c
m) are arranged in a square lattice with a pitch of 1.52 cm, and 7 × 7
A total of 9 fuel assemblies including one central fuel assembly 1a and eight fuel assemblies 1b surrounding the central fuel assembly 1a are arranged in a 3 × 3 array through water gaps 2 in two rows. There is. The fuel assembly 1b at the lower center of the nine fuel assemblies
The neutron flux monitor 3 is arranged along the line to measure the neutron flux.

Next, the operation of the above configuration will be described. FIG. 2 is a calibration curve diagram showing the correlation between the reciprocal of the thermal neutron flux and the fuel enrichment of the central fuel assembly in the plurality of fuel assemblies in which the fuel assemblies shown in FIG. 1 are arranged. 1
The fuel enrichment is separately given to each of the central fuel assembly 1a of the body and the eight fuel assemblies 1b arranged around this,
As a result of performing the two-dimensional three-group diffusion calculation under the condition of “with fixed neutron source”, the calculation result shown in FIG. 2 was obtained. Assuming that 0.1 n / s of neutrons are emitted per unit volume of UO ₂ as a neutron source, the obtained thermal neutron flux is in units of n / s / cm ² (absolute value) It is preferable to use the measured value for the rate and use the thermal neutron flux standardized by that value). From FIG. 2, when the fuel enrichment of the peripheral fuel assembly 1b is low, the reciprocal of the thermal neutron flux is large. That is, it is understood that the thermal neutron flux is small, and the thermal neutron flux is large as the fuel enrichment of the central fuel assembly 1a is large. Such characteristics are of course the same in the case of fast neutron flux and epithermal neutron flux.

The calibration curve diagram of FIG. 3, on the other hand, shows the results obtained by performing the well-known "eigenvalue mode" two-dimensional three-group diffusion calculation without the external neutron source. FIG. 2
The calculation with a fixed neutron source shown in is the "neutron flux mode"
It is sometimes called the calculation of neutron flux and obtains the absolute value of the neutron flux corresponding to a given external neutron source strength. This Figure 3
In the calculation of the "eigenvalue mode" of, the effective multiplication factor keff is obtained, and the neutron flux obtained at the same time is a virtual relative value.
"Neutron flux mode" and "eigenvalue mode", except for selecting the absolute value of the neutron flux or keff value,
Calculation is performed under the same calculation condition. From the viewpoint of ensuring criticality safety, this keff value is usually used, and in any case keff value
<1 must be secured. The fuel storage unit is usually designed so that the maximum keff value including various errors does not exceed 0.95. Therefore, it is considered that the actual keff value in the normal fuel assembly storage part is often 0.8 to 0.9. In the fuel storage unit for actually measuring the subcriticality (effective multiplication factor), calibration curve diagrams corresponding to FIGS. 2 and 3 are prepared in advance. Neutron source is Cm accumulated in irradiated fuel
-244, Cm-242, Pu-238, Pu-240, Pu-242, Am-241 etc. are neutrons emitted by spontaneous fission or (α, n) reaction with oxygen contained in fuel. . The neutron source intensity changes because they change with the cooling time, especially while Cm-242 (163 days) having a short half-life remains. Therefore, it is preferable to measure the spontaneous neutron source intensity and the distribution of the irradiated fuel assembly immediately before it is stored in the storage device. The measurement method has already been developed and is disclosed as an emission neutron measurement method, a neutron emission rate measurement method, or the like.

Next, one procedure for carrying out the present invention in an irradiation fuel assembly storage device will be described with reference to the layout view of the fuel assembly in FIG. First, the central fuel assembly 1a means a standard fuel assembly at a position where the reactivity effect is high.
Further, the peripheral fuel assembly 1b means an average fuel assembly around the standard fuel assembly. It is desirable that the irradiated fuel assemblies housed in the fuel housing section have as uniform characteristics as possible, which allows easy measurement work and high accuracy. New fuel separately for standard fuel assembly
(U235) may be used, but the neutron emission rate, burnup, fissile concentration, irradiation fuel assembly evaluated neutron multiplication factor using a neutron emission rate measurement method (for example, U235 + P239 +
P241 etc.) may be used. Next, in the calibration curve of the calibration curve diagram corresponding to the calibration curve diagram of FIG. 2, from the values of the fissile concentration and the thermal neutron flux of the standard fuel assembly (central fuel assembly 1a) of known composition, the effective of the peripheral fuel assembly 1b is determined. Of the peripheral fuel assembly 1b using the calibration curve of the calibration curve diagram corresponding to the calibration curve diagram of FIG.
The effective neutron multiplication factor (keff) in the fuel storage part is calculated from the effective fissile concentration (horizontal axis) and the fissile concentration of the central fuel assembly indicated by the parameter. Further, the calibration curve diagram of FIG. 4, which corresponds to FIG. 2 above, shows the effective fissile concentration of the peripheral fuel assembly as the horizontal axis and the fissile concentration of the central fuel as a parameter. Can be used similarly to FIG.

FIG. 5 is a plan view showing an example of a fuel basket in which a spent fuel assembly of a boiling water reactor (BWR) is housed in a cask. A spent fuel assembly housing in which 52 fuel assemblies are housed A fuel cell 11 of the fuel basket 10 which is a device is manufactured with a lattice material 12, and an outer cylinder 13 is provided on the outer circumference. A neutron monitor insertion hole 14 is arranged in a part between the outer cylinder 13 and the fuel cell 11. When practicing the present invention, the fuel basket 10 is submerged and the fuel assembly, not shown, is loaded or unloaded in water. In a preferred embodiment of the present invention, if the burnup value and its axial distribution, the fuel composition and its axial distribution, and the cooling time are relatively short (for example within 1.5 years),
Fuel assemblies having substantially the same cooling time are loaded in a predetermined procedure. Under these conditions, the characteristics of the fuel are uniform, and it is expected that the neutron emission rate is also uniform. Therefore, the measurement accuracy is high, and the measurement work and auxiliary calculation are easy. However, these conditions are not essential. This is because it is inevitable to use the result of the auxiliary calculation in implementing the present invention.

According to the standard procedure of the present invention, the neutron monitor 3 shown in FIG. 1 is inserted into at least one of the neutron monitor insertion holes 14, and the neutron monitor is sequentially loaded into the fuel cell 11 while the neutron monitor is being loaded. Monitor bundle level and fuel cell 11
Load all fuel assemblies. With this, after confirming that the fuel loading in the fuel basket 10 which is the spent fuel assembly storage device is sufficiently subcritical, the irradiated fuel assembly at the position where the reactivity effect is high is pulled out, and the standard fuel assembly is replaced instead. To load. In FIG. 5, the four irradiated fuel assemblies in the central portion marked with S are replaced with standard fuel assemblies.
Since the four irradiated fuel assemblies 1 are replaced one by one with the standard fuel assembly, the neutron flux level is constantly monitored by the neutron monitor 3, which causes the neutron flux level to be erroneously critical without realizing it. It does not happen and the criticality safety is secured. Fuel cell located in the center marked with S
The four irradiated fuel assemblies loaded in 11 are the positions where the reactivity effect is high, and the keff value is relatively large (keff:
Therefore, it is possible to effectively change the keff value and the neutron flux level in the fuel basket 10 by substituting a small number of standard fuel assemblies for measurement.

Although FIG. 5 shows an example in which four fuel assemblies are replaced in the central portion of the fuel basket 10, it is necessary to limit the number of fuel assemblies to four, and it is essential that the replacement positions are adjacent to each other. Instead, they may be dispersed to some extent and replaced. However, the outermost zone is generally not preferable. This is because the reactivity effect is small and is therefore disadvantageous for high-accuracy measurement, and if the substitution is performed in the vicinity of the neutron monitor 3, the local change in the neutron flux level is more remarkable than the keff value in the fuel basket 10. This is because Further, unlike the standard example described above, in the loading process, if the criticality safety after 100% loading by the irradiation fuel assemblies and the standard fuel assemblies is sufficiently guaranteed, the loading order need not be specified. Further, although the fuel storage device is shown as the case of the fuel basket 10 for the cask in the above-described embodiment, the present invention can be effectively applied to a fuel storage rack or the like installed in an underwater pool.

[0015]

As described above, according to the present invention, a part of the irradiated fuel assemblies is disposed at a position where the reactivity effect is high in the fuel assembly housing portion of the subcritical fuel assembly housing device housing the irradiated fuel assemblies. Measures the neutron flux when it is replaced with a standard fuel assembly of known composition, and evaluates the effective composition around the standard fuel assembly using a calibration curve obtained separately from the composition and neutron flux of the standard fuel assembly. However, from this composition and the composition of the standard fuel assembly, it is possible to easily know the neutron effective multiplication factor in the storage portion of the irradiated fuel assembly by using a separately obtained calibration curve. There is an effect that the safety can be secured quantitatively.

[Brief description of drawings]

FIG. 1 is a layout view of a fuel assembly showing an embodiment of the present invention.

FIG. 2 is a calibration curve diagram showing the correlation between the reciprocal of the thermal neutron flux and the fuel enrichment of the central fuel assembly in the fuel assembly arrangement of FIG.

FIG. 3 is a calibration curve diagram showing the correlation between the effective multiplication factor and the fuel enrichment of peripheral fuel assemblies in the fuel assembly arrangement of FIG.

FIG. 4 is a calibration curve diagram showing the correlation between the reciprocal of the thermal neutron flux and the fuel enrichment of the central fuel assembly in the fuel assembly arrangement of FIG.

FIG. 5 is a plan view of a fuel basket that stores a spent fuel assembly in a cask.

[Explanation of symbols]

1a ... central fuel assembly, 1b ... peripheral fuel assembly, 2 ... water gap, 3 ... neutron flux monitor, 10 ... fuel basket,
11 ... Fuel cell, 14 ... Neutron monitor insertion hole.

Claims (1)

[Claims] 1. A non-critical irradiation fuel storage unit in which a plurality of irradiation fuel assemblies are stored, and a part of the irradiation fuel assemblies at a place having a high reactivity effect is replaced with a standard fuel assembly having a known composition. The neutron flux in the standard fuel assembly is measured, and the effective composition around the standard fuel assembly is obtained using the calibration curve separately obtained from the composition of the standard fuel assembly and the neutron flux value. The neutron effective multiplication factor in the irradiation fuel storage, wherein the neutron effective multiplication factor of the fuel storage unit in which the irradiation fuel assembly is stored is obtained using a calibration curve separately obtained from the composition of the above.