JPS6316298A - Nondestructive measuring method of spent nuclear fuel aggregate - 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 [Object of the Invention] (Industrial Application Field) The present invention relates to a method for non-destructive measurement of spent fuel assemblies for evaluating the nuclear properties of spent fuel assemblies by non-destructive measurement.

(Prior art) Generally, spent fuel assemblies taken out from nuclear reactors are stored in fuel storage pools for a certain period of time to attenuate radioactivity with a short half-life, and then stored in transport containers and transported to reprocessing plants. or transported to long-term storage facilities.

When such a spent fuel assembly is found at a nuclear power plant, data such as cut-off enrichment and burnup are passed to the receiving side of the spent fuel assembly, such as a reprocessing plant or long-term storage facility. .

The sender of spent fuel assemblies from nuclear power plants, etc. carries out the transport work with as little error as possible, but when receiving spent fuel assemblies, the receiving side ensures that they are subcritical through various storage and processing steps. In order to maintain compliance, it is necessary to reconfirm that there are no errors by conducting independent measurements of the burnup, fissile nuclide concentration, etc. of spent fuel assemblies.

Conventionally, the so-called French method (H, DARROUZ) has been used as a method to evaluate the nuclear properties of such spent fuel assemblies.
ET, et, al, IAEA-260/20198
Two methods are known: 2) and West German method % formula % ARDA-2, P3961980).

(Problems to be solved by the invention) However, in such conventional methods, it is not possible to repeatedly measure important combustion parameters such as burnup and plutonium concentration using different methods, and there is a problem with the reliability of measured values. Ta.

Furthermore, since the West German method uses a neutron multiplication rate measurement method, an external neutron source such as Cf252 is required to measure the neutron multiplication rate, and this external neutron source must be updated at regular intervals due to its half-life. There was also the problem.

Therefore, the present invention is a method developed to solve these problems, and it greatly increases the reliability of measured values by repeatedly measuring important combustion parameters such as burnup and plutonium concentration using different methods. It is an object of the present invention to provide a non-destructive measurement method for spent fuel assemblies that can improve the performance of spent fuel assemblies and determine the concentration of fissile nuclides by measuring the neutron multiplication factor without using an external neutron source.

[Structure of the Invention] (Means for Solving the Problem) That is, the present invention includes a gamma ray spectrum analysis step for measuring and analyzing the spectrum of gamma rays extracted from a spent fuel assembly, and a gamma ray spectrum analysis step for measuring and analyzing the spectrum of gamma rays extracted from a spent fuel assembly. a spontaneous neutron emission rate measurement step of measuring the rate of spontaneous neutron emission caused by the spent fuel assembly; a spontaneous neutron multiplication rate measuring step of measuring the multiplication rate by changing the multiplication characteristic by which the spontaneous neutrons released from the spent fuel assembly are multiplied;
b) Determine the burnup, plutonium concentration, and cooling time through the gamma ray spectrum analysis step, and determine whether the spent fuel assembly uses uranium fuel or plutonium-enriched fuel, and (b) spend fuel assembly. The spent fuel assembly uses uranium fuel and the burnup is above a predetermined value, and the spent fuel assembly uses plutonium-enriched fuel, and the cooling time is longer than a predetermined value. Sometimes, the initial enrichment of the spent fuel assembly is determined by deriving the burnup and plutonium concentration through a spontaneous neutron emission rate measurement step; (c) On the other hand, the spent fuel assembly uses plutonium enriched fuel. or when the spent fuel assembly uses uranium fuel and the burnup determined in the gamma ray spectrum analysis step is above a predetermined value, the neutron multiplication factor is measured in the spontaneous neutron multiplication factor measurement step. The fissile nuclide concentration and infinite multiplication factor are derived from this neutron multiplication factor to obtain the initial enrichment of the spent fuel assembly.

(Function) The method for non-destructive measurement of spent fuel assemblies of the present invention combines gamma ray spectrum analysis, spontaneous neutron multiplication factor measurement, and spontaneous neutron multiplication factor measurement. The burnup (BtJ), plutonium 8 degrees (Pu/U), and cooling time (Tc) are determined, and it is determined whether the spent fuel assembly is made of uranium fuel or plutonium-enriched fuel, and the burnup (Bu/U) is determined. ) and cooling time (Tc>), the neutron emission rate (S), burnup (BU), and plutonium concentration (Pu/U) are determined by the spontaneous neutron emission rate measurement method, while the nuclear fission rate is determined by the spontaneous neutron multiplication rate measurement method. sexual nuclide concentration (Fiss) and effective multiplication factor (
keff) and infinite multiplication factor (kOO).

(Example) The details of the method of the present invention will be explained below with reference to one embodiment.

In this example, as reference data, the average initial enrichment of the fuel assemblies (εi) was obtained from the shipper of the spent fuel assemblies.
It is assumed that a spent fuel assembly is sent to the receiving side along with data such as burn-up (BtJ), date of completion of irradiation, etc.

FIG. 1 is a flowchart showing an embodiment of the method for non-destructive measurement of spent fuel assemblies of the present invention.

As shown in the figure, in this example, burnup (BUD) and plutonium concentration (
(often expressed as Pu/L concentration ratio), cooling time (Tc)
, and determine whether the spent fuel assembly uses uranium fuel or Plutoni-Kum enriched fuel. In this gamma ray spectrum analysis method, for example, as shown in FIGS. 2 to 4, the photopeak count rate ratio C5134
A calibration curve using /Cs137 and Pr144/Cs137 is used.

The value of the cooling time (Tc) obtained in this way is 2 to 2
．． For more than 5 years, the burn-up (BLI) value is 10-15GW
If d/ is greater than or equal to d/, the spontaneous neutron emission rate measurement method can be applied.

Additionally, by monitoring the size of the gamma ray photopeak emitted from Rh106, it is possible to determine whether the spent fuel assembly uses uranium fuel or plutonium-enriched fuel. . In the case of plutonium-enriched fuel, the spontaneous neutron emission rate measurement method can be applied even if the burnup (BU) value is low.

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

The spontaneous neutron emission rate measurement method uses a 51 calculation code whose validity has been confirmed to some extent by comparison with actual measured values.
Neutron emission rate 304 excluding the contribution of 42, or Cm
Neutron emission rate S4 and burnup (BUD, Pu
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 curves are used as a calibration curve.

In other words, using the fuel assembly average initial enrichment (εi) given for a large number of fuels, the burnup (BU) is determined from the neutron emission rate by the spontaneous neutron emission rate measurement method, and compared with the sender's data. Ru.

Further, the burnup (BU) ratio is determined for each fuel, and an average value of the ratios for a large number of fuels is created. And for example 15
Those with a significant difference of % 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 the spontaneous neutron emission rate measurement method is
) is used to determine the burnup (BtJ) of each fuel assembly.

Further, although the photopeak count value obtained by gamma ray spectroscopy is proportional to the burnup (BU), it may be difficult to obtain the proportionality coefficient by gamma ray spectroscopy. In such a case, the proportional coefficient for determining the burnup <BU) from the photopeak count value of C5137 determined by gamma ray spectroscopy should be replaced by the burnup (BU) obtained by the spontaneous neutron emission standard method. It can also be determined to match.

The total concentration of all Pu nuclides (Pu) can be obtained by both gamma ray spectroscopy and spontaneous neutron emission rate measurement, so the obtained results should be comprehensively compared to make them more reliable. be able to.

The average initial enrichment of each fuel assembly (ε
i) is the fuel assembly average initial enrichment (εi) at which the burnup (BU) determined by gamma ray spectroscopy and the burnup (BU) determined by spontaneous neutron emission rate measurement method match. 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.

The fissile nuclide concentration (FiSS) of one fuel assembly is determined by the relationship between the calculated fissile nuclide concentration (FiSS) and burnup (BU), or the fissile nuclide concentration (FiSS) determined by calculation.
It can be determined using a correlation curve between Pu ss) and the total a degree of all Pu nuclides (Pu). This fissile nuclide concentration (F
i SS) may be any of the total 'a degrees, uranium 2351 degrees, plutonium 239 permeability, plutonium 241 concentration, etc.

If the cooling time (Tc> is about 2 years or less or the burnup (
Btu) is less than 10 to 15 GWd/ (however, even lower burnup is possible for plutonium-enriched fuel), it becomes difficult to apply the spontaneous neutron emission rate measurement method.

On the other hand, the spontaneous neutron multiplication rate measurement method is unrelated to the length of the cooling time (TC>), as long as it is possible to measure the spontaneous neutron emission rate.For plutonium-enriched fuel, even if the burnup is small, it can be It is possible to measure the emitted neutrons, and this spontaneous neutron multiplication rate measurement method can be implemented.However, in the case of uranium fuel, measurement of the spontaneous neutron emission rate is not possible unless a burnup of at least 5 GWd/ has been achieved. Due to this difficulty, only the results of gamma ray spectroscopy can be obtained.Therefore, in this case, special treatment is usually given as important parameters cannot be measured using different methods.Such fuel assemblies include, for example, In the fuel assembly of a boiling water reactor (BWR), only the neutron poisoning effect of the burnable poison (Gd) has disappeared, and the U235 depletion has not progressed very much in a damaged fuel assembly. Such fuel assemblies should of course be treated specially from the viewpoint of criticality safety.

In the spontaneous neutron multiplication rate measurement method, the (α, n) difference between neutrons emitted from inside the spent fuel assembly by spontaneous fission and alpha rays emitted by the alpha decay of oxygen and transuranium nuclides contained in the fuel part is calculated. Since the neutrons released in the reaction become a neutron source (called spontaneous neutron source, primary medium/4 neutron source, intrinsic neutron source, internal neutron source, etc.), Cf252
There is no need to use an external neutron source (called an external neutron source because it is not a spontaneous neutron source specific to the fuel assembly even if it is placed inside the fuel assembly). There are several types of external neutron sources, but all have the problem of having to be updated at regular intervals due to their half-lives. For example, the half-life of the Cf252 neutron source, which is considered optimal as an external neutron source, is relatively short at 2.65 years, so it must be renewed every short period.

In the case of spontaneous neutron sources, there is no problem of updating the neutron source, except in the case of uranium fuel with a low burnup.

In the spontaneous neutron multiplication factor measurement method using spontaneous neutrons as a neutron source, the multiplication characteristics of the fuel assembly are changed by replacing at least part of the medium inside or outside the fuel assembly with a medium other than fuel. , is a method of determining the multiplication factor from the change in the neutron flux that changes thereby, and has already been disclosed in Japanese Patent Application Laid-Open No. 52-63595 and Japanese Patent Application Laid-open No. 52-63595.
3-52893@publication, LANL-9494-MS,
It is disclosed in EPRI-NP-2818.

Using the spontaneous neutron multiplication factor measurement method, for example, the effective multiplication factor (keff) when only one spent fuel assembly is placed underwater
) can also be determined, and the fissile nuclide concentration (F i ss) can also be determined. The infinite multiplication factor (koo) representing the unique characteristics of the fuel assembly is calculated from the infinite multiplication factor (koo) using a conversion factor (probability that neutrons will not leak) obtained by calculation. Values such as the infinite multiplication factor (kOo) and fissile nuclide concentration (Fiss) obtained by the spontaneous neutron multiplication rate measurement method are based on the correction of the multiplication effect necessary for the spontaneous neutron emission rate measurement method and the calibration curve obtained by calculation. It can also be used to evaluate the validity of values such as the above-mentioned infinite multiplication factor (koo) and fissile nuclide concentration (F-iss) derived semi-experimentally.

In the spontaneous neutron multiplication factor measurement method, the ratio (φ/φ0) of the neutron flux (φ0) before changing the system medium to the neutron flux (φ) after changing it is the effective multiplication factor (keff) or infinite multiplication factor. (koo) Utilizes the characteristics having the relationship shown in Figure 6. This correlation characteristic is usually obtained by calculation, but its validity can also be evaluated by experiment. When creating a correlation curve, the composition must be known or the effective multiplication factor (keff
) using known fuel assemblies.

That is, first, the effective multiplication factor (keff) and the fissile nuclide concentration (Fi ss) are determined based on the measured neutron flux ratio (φ/ψO). This effective multiplication factor (kef
The infinite multiplication factor (kOo) is determined by calculation based on f). By calculation, the correlation between the infinite multiplication factor (koO) and the neutron flux ratio (φ/φ0) was created as a calibration curve with the fuel assembly average initial enrichment (εi) as a parameter, as shown in Figure 8. The infinite multiplication factor (kC/Q) determined by this spontaneous neutron multiplication factor measurement method is
- and the burnup (8 tJ) determined by gamma ray spectroscopy, the fuel assembly average initial enrichment (εi) is determined. .

Fissile nuclide concentration (Fiss), i.e. uranium-235
The total concentration of plutonium 239 and plutonium 241 can be directly determined by spontaneous neutron multiplication factor measurement, but the concentration of each nuclide is determined using a calibration curve as shown in FIG. 7, for example, calculated in advance. In addition, the total concentration of Pu all nuclides (Pu) and the concentration of fissile nuclides (Pu) calculated in advance are calculated as follows:
S) is determined from a calibration curve using the average initial mm shrinkage (εi) of the fuel assembly as a parameter.

As described above, the non-destructive measurement of the spent fuel assembly is completed. Thereafter, by integrating the various data and comparing it with the data from the sender, the spent fuel assembly on the receiving side can be managed safely and reliably.

[Effects of the Invention] As described above, according to the method for non-destructive measurement of spent fuel assemblies of the present invention, gamma ray spectrum analysis, spontaneous neutron emission rate measurement, and spontaneous neutron multiplication rate measurement can be combined. This makes it possible to reliably perform non-destructive measurements of spent fuel assemblies regardless of the cooling time of the spent fuel assemblies, and to double measure especially important combustion parameters such as burnup and plutonium concentration. This greatly improves the reliability of measured values, and in general, it is possible to measure the neutron multiplication factor without using an external neutron source, which eliminates the need to update the external neutron source at regular intervals. There is also the effect of not having one.

[Brief explanation of the drawing]

FIG. 1 is a flowchart showing an embodiment of the method for non-destructive measurement of spent fuel assemblies of the present invention, and FIGS. 2 to 8 are graphs conceptually showing various correlation curves used in the present invention. Figure 2 Figure 3 Figure 4M 51st! J Figure 8

Claims (1)

[Claims] (1) A gamma ray spectrum analysis step of measuring and analyzing the spectrum of gamma rays emitted from the spent fuel assembly, and spontaneous neutron emission rate measurement of measuring the emission rate of spontaneous neutrons emitted from the spent fuel assembly. and multiplying the spontaneous neutrons emitted from the spent fuel assembly by replacing at least a portion of at least one of the neutron moderating medium outside or inside the spent fuel assembly with a material having different nuclear properties. a spontaneous neutron multiplication rate measurement step of measuring the multiplication rate by changing the multiplication characteristics
(b) determining the burnup, plutonium concentration, and cooling time through the gamma ray spectrum analysis step; and determining whether the spent fuel assembly uses uranium fuel or plutonium-enriched fuel; The spent fuel assembly uses uranium fuel and the burnup is above a predetermined certain value, and the spent fuel assembly uses plutonium-enriched fuel and the burnup is above a predetermined certain value. (c) when the spent fuel assembly has the cooling time of or when the spent fuel assembly uses uranium fuel and the burnup determined in the gamma ray spectrum analysis step is equal to or higher than a predetermined value. A spent fuel characterized in that a neutron multiplication factor is determined by a spontaneous neutron multiplication factor measurement step, and a fissile nuclide concentration and an infinite multiplication factor are derived from this neutron multiplication factor to determine the initial enrichment of the spent fuel assembly. Non-destructive measurement method for aggregates.