JP2635860B2 - Radioactivity evaluation method for solidified radioactive waste - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for evaluating the radioactivity of a solidified radioactive waste, and more particularly to a high-level radioactive solidified material generated in the reprocessing of nuclear fuel materials used in nuclear power plants. The present invention relates to a radioactivity evaluation method for solidified radioactive waste, in which the absolute value of radioactivity of radionuclides is determined by nondestructive measurement.

[0002]

2. Description of the Related Art Generally, it is important from the viewpoint of safety evaluation and the like to measure and evaluate the radioactivity of a solidified radioactive waste handled in a nuclear fuel cycle facility or the like.

[0003] As one of such solidified radioactive wastes, there is a high-level radioactive solidified solid generated in the reprocessing of nuclear fuel materials used in nuclear power plants. However, there is no conventional technology for non-destructive measurement of the radioactivity of radioactive nuclides in such high-level radioactive vitrified solids.Before vitrification, directly measure γ-rays, α-ray neutrons, etc. By doing so, there is only a method of quantifying the radioactivity of the nuclide, or a method of measuring only the radiation dose rate of the vitrified material using a dose rate measuring device.
In addition, there is a test measurement example of a small test solidified glass containing RI with a known concentration.

[0004]

As described above, conventionally, prior to vitrification, direct γ-rays, α-rays,
Although it is possible to quantify radioactivity by measuring neutron beams, etc., in order to quantify radioactivity of vitrified radioactive waste, it must be destroyed again and analyzed for radioactivity. In practice, it cannot be performed in terms of work efficiency and worker exposure control.

[0005] In the measurement of the radiation dose rate as described above, the absolute value of the radionuclide cannot be determined. For this reason, in the disposal of radioactive waste, it is preferable to grasp the radioactivity or radioactivity concentration for each radionuclide for safety evaluation after disposal, but such evaluation has not been performed so far.

The present invention has been made in view of such conventional circumstances, and the radioactivity of each radionuclide contained in the solidified radioactive waste can be easily quantified in absolute value by nondestructive measurement. An object of the present invention is to provide a method for evaluating the radioactivity of a solidified radioactive waste, which can improve the safety as compared with the related art.

[0007]

That is, the present invention is configured such that an RI source having a known radioactivity concentration can be disposed therein, and that the position of the RI source can be moved to a plurality of locations. The step of measuring emission γ-rays while rotating the simulation sample for radioactivity calibration and obtaining the detection efficiency of γ-rays for radioactive nuclides comprises the step of measuring the RI in the simulation sample for radioactivity calibration.
The detection efficiency distribution is repeatedly obtained while changing the position of the radiation source to derive an average radioactivity detection efficiency per unit volume from the detection efficiency distribution, and thereafter, γ-rays of the solidified radioactive waste to be evaluated are obtained. , And the measurement result is corrected by the average radioactivity detection efficiency per unit volume.

[0008]

According to the radioactivity evaluation method for a solidified radioactive waste of the present invention having the above-described configuration, an RI source having a known radioactivity concentration can be disposed inside the source, and the position of the RI source can be determined at a plurality of locations. Use a simulated specimen for calibration of radioactivity that is configured to be movable. That is, Cs-137 or Cos-137 is used as a simulated radiation source for radioactivity calibration so that the detection efficiency of gamma rays emitted from a vitrified radiation source can be measured three-dimensionally.
Using a simulated vitrified body capable of distributing and arranging RI radiation sources such as -60, radiation from the simulated vitrified body by changing the arrangement of the RI source under the same measurement conditions as the radioactivity measurement of the high-level radioactive vitrified body By repeating the measurement of the radioactivity, the average detection efficiency of radioactivity per unit volume of the vitrified product is determined.

That is, for example, first, an RI ray source is arranged at an arbitrary radial position of the simulated vitrified body for calibration, and the measurement of the detection efficiency at that position is performed while the measurement position in the height direction is fixed. By repeatedly changing the position of the source in the radial direction, the distribution of detection efficiency in the radial direction is obtained. Next, the measurement efficiency distribution in the radial direction can be obtained by repeatedly measuring the measurement position in the height direction with the measurement position in the height direction being finely changed to obtain the detection efficiency distribution in the height direction. An average detection efficiency per unit volume in consideration of the distribution of the detection efficiencies can be obtained by obtaining the optimum detection efficiency distribution and averaging the entire distribution.

The detection efficiency and the γ of the nuclide to be evaluated
The radioactivity of the radionuclide contained in the entire high-level radioactive vitrified matter is quantified by correcting the count value obtained by measuring the gamma ray spectrum from the outside of the high-level radioactive vitrified substance using the radiation rate of the radiation.

By using this method, it is possible to calibrate radioactivity measurement without using, for example, a large-sized high-level radioactive solidified body which is difficult to manufacture and handle and having the same shape as a calibrated glass solidified body. In order to determine the average detection efficiency, measurement of high-level radioactive glass
It is only necessary to measure the gamma-ray spectrum from an external portion at any position.

[0012]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The details of the radioactivity evaluation method for solidified radioactive waste of the present invention will be described below with reference to the drawings.

FIG. 1 shows a configuration of a radioactivity measuring system according to an embodiment of the present invention. In FIG. 1, reference numeral 1 denotes a high-level radioactive vitrified body. The high-level radioactive vitrified body 1 is mounted on a vitrified body driving device 2 configured to be vertically movable and rotatable, and γ-rays emitted from the high-level radioactive vitrified body 1 are shielded from radiation. The radiation detector 5 is configured to count through a γ-ray collimator 4 installed so as to penetrate the wall 3.

On the other hand, in FIG. 2, reference numeral 6 denotes a simulated vitrified body for absolute value calibration loaded with an RI source. The simulated vitrified body 6 is made of a material similar to that of the high-level radioactive vitrified body 1 to be evaluated, and is formed in a columnar shape having substantially the same diameter. It is shorter and smaller than the body 1. Further, inside the simulated vitrified body 6, a desired RI radiation source is configured to be loaded, and the RI
It is configured such that the radiation source can be arranged at a plurality of different positions in the radial direction. In this example, as shown in FIG. 5 described later, from the center of the simulated vitrified body 6,
It is configured so that RI radiation sources can be arranged at eleven locations up to the radial outer periphery.

In this embodiment, prior to the measurement of the radioactivity of the high-level radioactive vitrified body 1, first, the simulated vitrified body 6 is placed on the vitrified body driving device 2, and the simulated vitrified body 6 is rotated. While performing the measurement, the radiation detector 5 measures the radiation.

FIG. 3 shows the procedure of radioactivity measurement and evaluation in the present embodiment. For example, when radioactivity of Cs-137 in a high-level radioactive vitrified substance is evaluated, the evaluation is performed according to the following procedure. Do.

That is, the high-level radioactive vitrified product 1
Prior to the measurement of the radioactivity, a simulated vitrified body 6 in which a Cs-137 source having a known concentration was loaded at an arbitrary radial position.
(100).

In this measurement, first, the simulated vitrified body 6 is placed on the vitrified body driving device 2 and set at a certain height position while rotating the simulated vitrified body 6 while detecting the radiation detector. The rotational average count rate at a certain height position is measured by performing γ-ray spectrum measurement in step 5 (101).

Next, the simulated vitrified body 6 is slightly moved in the height direction (102), and the rotational average count rate is measured again (101).

By repeating the movement in the height direction and the rotation measurement, a distribution in the height direction of the count rate at an arbitrary radial position as shown in FIG. 4 is obtained (103). In addition,
The graph of FIG. 4 plots the counting rate at the height position with respect to the height position based on the center of the γ-ray collimator.

Next, the average count rate in the height direction is obtained from the area of the hatched portion in FIG.
By dividing the γ-ray generation amount per unit time of the s-137 source by the area that can be challenged by the γ-ray collimator at that radial position, the detection efficiency Eh per unit area at an arbitrary radial position is obtained ( 104).

Thereafter, the position of the Cs-137 source in the radial direction is changed (105), and the measurement and analysis of 101 to 104 are repeated to obtain a radial detection efficiency distribution as shown in FIG. 5 (106). .

The graph of FIG. 5 plots the value obtained by multiplying the detection efficiency Eh at each radial distance by the weight of the volume in the circumferential direction with respect to the radial distance in the center of the circle of the simulated vitrified body 6. Things. Then, an average detection efficiency E per unit volume for γ-rays emitted from Cs-137 in the simulated vitrified solid 6 is determined from the area of the hatched portion (107).

Next, as shown in FIG. 1, γ-ray spectrum measurement of the high-level radioactive vitrified product 1 is performed (2).
00).

That is, first, a γ-ray vitrified unit volume of γ-ray having an energy of 662 KeV emitted from Cs-137 is measured by a γ-ray spectrum measurement of the high-level radioactive vitrified body 1 at an arbitrary rest position in the height direction. The counting rate C per unit is measured (201). Then, it is assumed that the radioactivity in the vitrified material is homogeneously distributed (202), and this count rate C is divided by the γ-ray emission rate B of 662 KeV of Cs-137 (203). By dividing by the detection efficiency E obtained from the measurement of the simulated vitrified material 6, Cs- contained in one unit volume of the high-level radioactive vitrified material
137 radioactivity Av is obtained (204).

The radioactivity Av thus obtained is
Is multiplied by the volume V of the vitrified product (20
5), Cs-137 in the high-level radioactive vitrified product 1
Is determined (206).

As described above, in this embodiment, by using the simulated vitrified body 6 for simple calibration, the radioactivity can be evaluated by a simple calibration work. In addition, radioactivity evaluation can be performed for other nuclides in the same manner as in the case of Cs-137 described above.

Next, a case of evaluating the radioactivity of an arbitrary radionuclide contained in the high-level radioactive vitrified product 1 will be described with reference to FIG.

First, in the γ-ray spectrum measurement (300) of the simulated vitrified body 6 loaded with an RI ray source that emits γ-rays of energy ε of known concentration, the measurement is performed according to the above-described procedure, and γ-rays of energy ε are measured. The average detection efficiency E (ε) per unit volume of the vitrified body with respect to is determined (301).

Next, an RI source having a different energy ε is changed (302), and the detection efficiency E (ε) is similarly increased (3).
02). Such a process is repeated a plurality of times, and a detection efficiency function f (ε) is obtained from the obtained plurality of E (ε) (303).

On the other hand, γ of the high level radioactive vitrified material 1
In the line spectrum measurement (400), first, the counting rate C (ε) of an arbitrary nuclide is determined by gamma ray spectrum measurement (401), and the detection efficiency corresponding to the γ-ray energy of the nuclide is determined by a detection efficiency function f (ε). And the radioactivity Av per unit volume of an arbitrary nuclide contained in 1 in the high-level radioactive vitrified material in the same manner as in the above-described embodiment.
(Ε) is obtained (402).

Then, the radioactivity A (ε) of an arbitrary nuclide is determined from the volume of the vitrified material (403).

In this embodiment, in order to make the detection efficiency a function, measurement work of the detection efficiency is performed at a facility other than a facility having relatively low workability due to its structure and management system, such as a facility handling high-level radioactive waste. be able to. Therefore, the workability is improved, and the efficiency of the radioactivity evaluation work is improved.

Next, another embodiment will be described with reference to FIGS.

FIG. 7 shows a scan measurement range of the high-level radioactive vitrified body 1. In this embodiment, the scanning range in the height direction of the high-level radioactive vitrified body 1 is the high-level radioactive vitrified body. Set to a range longer than the height of 1. Therefore, for example, 662 KeV of Cs-137
The count of γ-rays is as shown by the shaded portion 30 in FIG.

In this embodiment, as shown in FIG. 8, in the measurement of the γ-ray spectrum of the high-level radioactive vitrified product 1 (500), first, the average count rate Cave within the scan range is obtained.
Is obtained as a value obtained by dividing the area of the shaded portion 30 by the scan time T (501), and this Cave 30 is calculated as Cs-1.
37 divided by the detection efficiency E per unit volume (50
2) By dividing by γ-ray emission rate B of Cs-137 (503), radioactivity Av of Cs-137 contained per unit volume of the high-level radioactive vitrified product is obtained (5).
04).

The radioactivity Av per unit volume
Multiplied by the scan volume Vsca (505),
The radioactivity A of Cs-137 in the high-level radioactive vitrified matter is determined (506).

According to this embodiment, the absolute value of radioactivity can be determined even when the volume of the high-level radioactive vitrified product 1 is unknown.

[0039]

As described above, according to the present invention,
The absolute value of the radioactivity of each radionuclide contained in the solidified radioactive waste can be easily determined by nondestructive measurement, and safety can be improved as compared with the conventional method.

[Brief description of the drawings]

FIG. 1 is a diagram showing a configuration of a measurement system according to an embodiment of the present invention.

FIG. 2 is a diagram showing a configuration of a measurement system according to an embodiment of the present invention.

FIG. 3 is a diagram showing a procedure for measuring and evaluating radioactivity in one embodiment of the present invention.

FIG. 4 is a diagram showing a count rate distribution in radioactivity calibration according to one embodiment of the present invention.

FIG. 5 is a diagram showing a detection efficiency distribution in radioactivity calibration according to one embodiment of the present invention.

FIG. 6 is a diagram showing a radioactivity measurement evaluation procedure in another embodiment.

FIG. 7 is a diagram for explaining a scan range in another embodiment.

FIG. 8 is a diagram showing a radioactivity measurement evaluation procedure in another example.

[Explanation of symbols]

 REFERENCE SIGNS LIST 1 high-level radioactive vitrified substance 2 vitrified substance driving device 3 radiation shielding wall 4 γ-ray collimator 5 radiation detector 6 simulated vitrified substance for calibration

Claims (1)

(57) [Claims] A simulated specimen for calibration of radioactivity in which an RI source having a known radioactivity concentration can be arranged and the position of the RI source can be moved to a plurality of locations. Measuring the emission γ-rays while performing the process of obtaining the detection efficiency of γ-rays for radioactive nuclides, while repeatedly changing the position of the RI source in the simulation sample for radioactivity calibration to obtain the detection efficiency distribution The average radioactivity detection efficiency per unit volume is derived from the detection efficiency distribution, and thereafter, the γ-ray of the radioactive waste solid to be evaluated is measured, and the measurement result is averaged per unit volume. A method for evaluating the radioactivity of a solidified radioactive waste, wherein the radioactive waste is corrected by a specific radioactivity detection efficiency.