JPS61139795A - Nondestructive measuring method of spent fuel - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention 1] The present invention relates to a method for non-destructive measurement of spent fuel for evaluating the nuclear properties of spent fuel by non-destructive measurement.

[Technical background of the invention and its problems] Generally, spent fuel taken out from a nuclear reactor is stored in a fuel storage pool for a certain period of time to attenuate its radioactivity, which has a short half-life, and then is stored in a transport container. Transported to reprocessing plants or long-term storage facilities.

When such spent fuel is transported from a nuclear power plant, data such as initial enrichment and burnup are passed to the spent fuel receiving side, such as a reprocessing plant or long-term storage facility.

Shippers of spent fuel from nuclear power plants, etc., carry out the removal work to be as error-free as possible, but those who receive the spent fuel must ensure that subcriticality is maintained throughout the storage and processing processes. It is necessary to reconfirm that there are no errors by conducting independent measurements of the initial enrichment, burnup, etc. of the spent fuel.

As a method to evaluate the nuclear properties of such spent fuel,
Conventionally, the so-called French method (M, DARROLJZET
, et at, IAEA-260/201982) and the West German system (H, WUERZ, IAEA-260/301)
982 and G, 5CHULZE, ESARDA-2,
P396 1980) are known.

The French method uses gamma ray spectrometry (GS), which measures the spectrum of gamma rays emitted from spent fuel.
This method evaluates the plutonium concentration by measuring the burnup and cooling time using a passive neutron measurement method (PN method), which measures spontaneous neutrons emitted from spent fuel.

However, in this French method, passive neutron measurement cannot be used when uranium fuel has a low burnup and cooling time is short, and gamma ray spectroscopy is used, so reactor operation It has the disadvantage of requiring a detailed history.

The West German method consists of the above-mentioned passive neutron measurement method and the active neutron measurement method (AN method), which places a neutron source on the side or inside the spent fuel and measures the neutrons multiplied by the neutrons emitted from this neutron source. This is a method to evaluate burnup, plutonium concentration, fission nuclide concentration, initial enrichment, neutron multiplication factor, etc. using

However, this West German method has the disadvantage that passive neutron measurement cannot be used for uranium fuel if the cooling time is short, and the cooling time cannot be evaluated.

In addition, as a method for measuring using passive neutron measurement when the cooling time is short, the present inventors have proposed
There is a method disclosed in No. 3-22993, but this method involves measuring neutrons twice with different cooling times, so passive neutron measurements are performed twice at least 2 to 3 months apart. There is a problem of poor workability.

[Object of the invention] The present invention has been made in response to such conventional circumstances,
The object of the present invention is to provide a method for non-destructively measuring spent fuel that can reliably measure spent fuel non-destructively regardless of the cooling time of the spent fuel.

[Summary of the Invention] That is, the present invention provides a gamma ray spectrometry method for measuring the spectrum of gamma rays emitted from spent fuel;
There is a passive neutron measurement method that measures spontaneous neutrons emitted from spent fuel, and an active method that places a neutron source on the side or inside the spent fuel and measures the neutrons multiplied by the neutrons emitted from this neutron source. The burn-up and cooling time of the spent fuel are determined using the gamma ray spectrometry method, and when the cooling time is equal to or greater than a predetermined value, the burn-up is determined using the passive neutron measurement method. Deriving the degree and plutonium concentration,
Thereby, the initial enrichment of the spent fuel is determined, and when the cooling time determined by the gamma ray spectrometry is equal to or less than the predetermined value, multiplied neutrons are measured by the active neutron measurement method. This is a method for non-destructive measurement of spent fuel, characterized in that the initial enrichment of the spent fuel is determined by deriving at least one of the fissile nuclide concentration or the infinite multiplication factor from the multiplied neutrons.

[Embodiment 1 of the invention The details of the method of the present invention will be explained below with reference to one embodiment.

In this example, the spent fuel is sent from the spent fuel shipping side to the receiving side along with data such as the fuel assembly average initial enrichment (εi), burnup (BU), and irradiation end date. This is assumed to be the case.

FIG. 1 is a flowchart showing an embodiment of the method for non-destructive measurement of spent fuel according to the present invention. As shown in the figure, in this example, the approximate cooling time (Tc) and burnup (BU) are first determined by gamma ray spectroscopy. In this gamma ray spectrum measurement method, for example, the second
As shown in Figure 4, the photobeak count rate ratio C6
A calibration curve using Pr 134/C8137, Pr 144/Cs 137 is used.

The value of the cooling time (Tc) obtained in this way is 2~
Over 2.5 years, burnup (BU) value is 10-15GW
Passive neutron measurement can be applied if d/ is greater than or equal to d.

Note that by monitoring the magnitude of the gamma ray photopeak emitted from Rh106, it is possible to determine whether the spent fuel is uranium fuel or plutonium fuel. In addition, in the case of plutonium fuel, the burnup (B
Even if the value of U) is low, the passive neutron measurement method can be applied.

Normally received spent fuel satisfies this condition, so by comparing it with the data sent from the shipping side, it is possible to confirm that there are no major errors.

In the passive neutron measurement method, Cm24 is calculated using a calculation code whose validity has been confirmed to some extent by comparison with actual measured values.
Neutron generation rate SO4 excluding the contribution of 2, or Cu
Neutron generation rate S and burnup (BU) from +244, P
A correlation curve with the total concentration of all nuclides (Pu), etc. is created in advance as shown in FIG. 5 using the fuel assembly average initial enrichment (ε1) as a parameter, and these are used as a calibration curve.

That is, using the fuel assembly average initial enrichment (εi) given to a large number of fuels, the burnup (BU) is determined from the neutron generation rate by the passive neutron measurement method and compared with the sender's data. Also, burnup (BU) for each body
The ratio is determined and an average value of the ratio is created for a number of fuels. For example, those with a significant difference of 15% or more are excluded.

This average value is used as a bias value for the calibration curve calculated in advance, and the calibration curve is thereby corrected.

Although it is difficult to accurately determine the burnup (BU) of each fuel assembly in fuel combustion management at the power plant, the total output of many fuel assemblies can be determined accurately through electrical output. Therefore, the average value of the ratio determined for a large number of fuel assemblies as described above is extremely reliable.

In this way, the burnup (BU) of passive neutron measurement method
The burnup (BU) of each fuel assembly is determined using the modified calibration curve for the fuel assembly, and the proportionality coefficient that determines the burnup (BU) is determined from the C3137 photobeak count determined by gamma ray spectrometry.

Note that although the photobeak count value is proportional to the burnup (BU), it is extremely troublesome to determine the proportionality coefficient by gamma ray spectroscopy. In addition, since the total concentration of all Pu nuclides (Pu ) can be obtained by both gamma ray spectrometry and passive neutron measurement, the obtained results should be comprehensively compared to obtain higher reliability. be able to.

The average initial enrichment of the fuel assembly for each fuel assembly (
ε1) is the fuel assembly average initial enrichment (εI) at which the burnup (BU) determined by gamma-ray spectrometry and the burnup (BU) determined by passive neutron measurement coincide. determined from a comparison of Since the types of fuel assembly average initial enrichment (εi) are usually very limited, they can be easily identified and determined.

Concentration of fissile nuclides in one fuel assembly (1” 1sS
) is the relationship between the calculated fissile nuclide concentration (Fils) and the burnup (BU), or the fissile nuclide concentration (Fils).
ss) and the total concentration of all nuclides (Pu). This fissile nuclide concentration (F
iss), the total concentration is uranium 235+11 degrees,
Either plutonium 23911 degrees, plutonium 241 m degrees, etc. may be used.

If the cooling time (Tc) is less than 2 years or the burnup (
When the BU) is less than 10 to 15 GWd/(excluding plutonium fuel), it becomes difficult to apply the passive neutron measurement method. Therefore, in this case, active neutron measurement is used.

This active neutron measurement method is a method in which a neutron source is placed on the side or inner surface of a fuel assembly, and the multiplied neutron flux φ due to the arrangement of the neutron source is measured on the side or inner surface of the fuel assembly.

This multiplied neutron flux φ is determined by the fissile nuclide concentration (Fiss) of the fuel assembly or the effective multiplication factor (k
There is an extremely close correlation with the fissile nuclide concentration (Fiss) or the effective multiplication factor (kerf) using this characteristic.

Although this active neutron measurement method can be applied regardless of the cooling time (Tc), it is slightly more troublesome to implement than the passive neutron measurement method, so it is not recommended in practice when the reliability of the passive neutron measurement method decreases or when the importance of the passive neutron measurement method decreases. It is preferable to apply it when it is high.

In this active neutron measurement method, a calibration curve can be created in advance by measurements using a standard fuel assembly whose composition is known.

That is, first, based on the measured multiplied neutron flux φ, the effective multiplication factor (ken) and the fissile nuclide concentration (Fiss
) is determined. The infinite multiplication factor (k output) is determined by calculation based on this effective multiplication factor (kefr).

Furthermore, through calculations, the correlation between the burnup (BU) and the multiplied neutron flux φ was created in advance as a calibration curve with the fuel assembly average initial enrichment (ε1) as a parameter, as shown in Figure 8. Using the infinite multiplication factor (koo) obtained by this active neutron measurement method and the burnup (8U) obtained by gamma ray spectrometry, the average initial enrichment (ε1) of the fuel assembly is calculated. It is determined.

Fissile nuclide concentration (Fiss), i.e. uranium-235
Although the total concentration of plutonium 239 and plutonium 241 can be determined directly by active neutron measurement, the concentration of each nuclide is calculated using a calibration curve shown in FIG. 7, for example. In addition, the total concentration of all PU nuclides (Pu) is calculated using the fuel assembly average initial enrichment (εl) between the total concentration of all PU nuclides (PU) and the fissile nuclide concentration (1:iss>), which are calculated in advance. determined by the calibration curve.

As described above, the non-destructive measurement of spent fuel is completed. Thereafter, by integrating various data and comparing it with the data from the sender, it is possible to ensure the safe and reliable management of spent fuel on the receiving side.

[Effects of the Invention] As described above, according to the method for non-destructive measurement of spent fuel of the present invention, spent fuel can be measured by combining gamma ray spectrometry, passive neutron measurement and active neutron measurement. Non-destructive measurement of spent fuel can be reliably performed regardless of the magnitude of the cooling time.

[Brief explanation of the drawing]

FIG. 1 is a flowchart showing an embodiment of the method for non-destructive measurement of spent fuel according to the present invention, and FIGS. 2 to 8 are graphs showing the concepts of various correlation curves used in the present invention. Representative Patent Attorney Satoshi Suyama - Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Figure 6 Figure 7 Figure 8

Claims (1)

[Claims] (1) Gamma ray spectrometry method that measures the spectrum of gamma rays emitted from spent fuel, passive neutron measurement method that measures spontaneous neutrons emitted from spent fuel, and neutron source on the side or inside of spent fuel. The active neutron measurement method measures the neutrons multiplied by the neutrons emitted from the neutron source, the burnup and cooling time of the spent fuel are determined by the gamma ray spectrometry method, When the cooling time exceeds a predetermined value, the burnup and plutonium concentration are derived by the passive neutron measurement method to determine the initial enrichment of the spent fuel, while the gamma ray spectrometry method determines the burnup and plutonium concentration. When the cooling time is equal to or less than the predetermined certain value, multiplied neutrons are measured by the active neutron measurement method, and at least one of the fissile nuclide concentration or the infinite multiplication factor is derived from the multiplied neutrons, and the used A method for non-destructive measurement of spent fuel, characterized by determining the initial enrichment of the fuel.