JPS6070385A - Discrimination of radioactivity in human body - 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] [Field of Application of the Invention] The present invention relates to a method for measuring radioactivity, and particularly to a method for measuring radioactivity suitable for identifying cases where body surface contamination and internal contamination occur simultaneously. be.

[Background of the invention]

Conventionally, the amount of radioactive substances ingested into the human body has been measured using bioassays or human counters (whole body counters).

The former method estimates the total amount of radioactive substances in the body from the amount of radioactive substances contained in exhaled breath or urine. The latter is a method of measuring the total amount of radioactive substances present in the subject using a plurality of radiation detectors placed near the outside of the body. All of these methods measure the total amount of radioactive material ingested into the body at the time of measurement. In either case, it is a method of estimating or measuring the total amount of radioactive material ingested into the body at the time of measurement. These methods cannot identify areas of radioactive contamination.

[Purpose of the invention]

An object of the present invention is to provide a method for measuring human body radioactivity that accurately identifies complex contamination when internal contamination and body surface contamination occur simultaneously.

[Summary of the invention]

A feature of the present invention is that the energy spectrum of gamma rays emitted from radioactive substances is measured, and the body axis (height) The purpose of this method is to identify complex contamination within the body and on the body surface by comparing the ratios of the photoelectric effect components and the Copton scattering components obtained by measuring the subject from two different directions in a plane perpendicular to the direction.

[Embodiments of the invention]

DESCRIPTION OF THE PREFERRED EMBODIMENTS A radioactivity measuring device which is a preferred embodiment of the present invention will be described with reference to FIGS. 1 and 2.

A radioactivity measuring device 2iI'i, a support frame 33 on which a bed 38 on which the person to be measured lies on his/her back, a driving section 4, an operating section 7,
It consists of a cutting section 10, a calculation section 14, and a display section 29. The drive section 4 includes a radiation detector 3, a drive device t5, a position detector 6, and a moving cart 37. The moving cart 37 is attached to the guide rail 32 via the support frame 33,
It moves along the guide rail 32. A drive device 5 is installed on a moving trolley 37. The guide rail 32 is provided with a rank (not shown), and a pinion (not shown) provided on the drive device 5 meshes with the rack.

The drive device 5 is provided with a motor (not shown), and the motor is connected to the pinion via a reduction gear. The guide rail 32 also supports a rotating arm 100 that rotates 360 degrees.
is supported by The rotary arm 100 includes a gear 101 and a geared reduction motor 10 for rotation in a plane perpendicular to the body axis.
2 are connected to each other, and by driving the deceleration motor 102, the radiation detector 3 can be moved to a position 103 in a plane perpendicular to the body axis of the subject.

The position of the radiation detector 3 is measured by a position detector 6 provided on the movable cart 37 and a position detector 104 (which detects the rotation angle of the rotary arm 1000) connected to the gear of the rotary arm 100. The support frame 33 is surrounded by a radiation shield (not shown).

The operation section 7 has an operation panel 8 and a monitor television 9. The operation panel 8 transmits an operation command to the drive device 5 and moves the drive device 5 along the guide rail 32 at a predetermined position. The monitor television 9 optically displays the output signal of the position detector 6.104 that detects the position of the radiation detector 3 and the operating status of the mobile trolley 37 and the like.

The al/measurement section 14 includes a radiation detector 3, a high voltage power supply 11, an amplifier 12, and a pulse height analyzer 13.

The high voltage power supply 11 applies a high voltage to the radiation detector 3 in order to operate the radiation detector 3. The radiation detector 3 is
Covered? Radioactivity of ill judge 1? To detect. Radiation detector 3
The output signal of is amplified by an amplifier 12. The amplifier 12 is connected to the pulse height analyzer 13'J? at connectors 34 and 39. and data acquisition interface 15. The wave height analyzer 13 calculates the wave market value of the γ-ray energy spectrum of the input signal.

(The Jt'i1 section 14 has a data acquisition interface 15, a medium heat processing device 16, and an external storage device 17. is transmitted to the central processing unit 16ic.External storage device 1
7 is ROM f ReadOnly Memory
) 18, RAM (RandomAccess Me
mory) 19, magnetic tape 20 and disk 2
Consists of 2. Address bus 27 and data bus 28
is connected to the central processing unit 16. ROMI 8. R
AMI 9. Magnetic tape interface 21. Disk interface 23゜Panel interface 24
．． The line printer interface 25 and the display memory 26 are connected to the 7 address bus 27 and the data node (228), respectively. The magnetic tape 20 is connected to the magnetic chip/surface 21, and the disk 22 is connected to the disk interface 23.

H,0M18 is a program that determines radionuclides from photoelectric peak energy, a program that calculates indicators of γ-ray energy spectrum, a program that identifies contamination, and a program that determines body surface contamination and internal contamination and calculates internal exposure dose, which will be described later. A calculation program such as a program for the radioactivity measurement method of this embodiment is stored. The RAM 19 stores numerical data necessary for calculations such as data for radionuclide determination and data for contamination identification, and also secures a work area. The operation panel 8 of the operation section 7 is
It is connected to the panel innoter face 24.

The display section 29 includes a line printer 30 and a display 31. Line printer 30 and display 31 are connected to line printer interface 25 and display memory 26, respectively.

The measurement of the radioactivity of the person to be measured 1 by the radioactivity measuring device 2, which is performed at t'i# as described above, will be explained below based on FIGS. 3 and 4. The person to be measured 1 is placed on the bed 38. The operator presses the first start button for radioactivity measurement on the operation panel 8. This command is transmitted to the central processing unit 16 via the panel interface 24. The central processing unit 16 also calls the entire program shown in FIG. 3 stored in the 0M18, and performs operations and calculations necessary for measuring the radioactivity of the person to be measured 1 according to the program. The central processing unit 16 sends a command to the drive device 5 via the panel innoter face 24 and the operation panel 8 to move the movable cart 37ft along the guide rail 32 (step 40). At the same time, a signal is sent to the connector 34 to connect the amplifier 12 and the contact 36 (step 41). Connector 39 is disconnected from contact 35. The radiation detector 3 measures radioactivity in the height direction of the person to be measured l as the movable trolley 37 moves. The output signal of the radiation detector 3 is amplified by the amplifier 12 and then input to the central processing unit 16 via the data acquisition interface 15 (step 42).
, are stored on magnetic tape 20 or disk 22. When the radioactivity measurement is completed, the radioactivity level distribution in the height direction of the person to be measured I is determined (step 43). The results are as shown in FIG. 5. At the position contaminated with 0 radioactive substances displayed on the display 31, a peak appears at the (Y) portion as shown in FIG. Multiple peaks will result if there is contamination at multiple locations. Since the person to be measured 1 is being examined in a room made of a radiation shield as described above, the measurement results are hardly affected by external radioactivity.

The effects of external radioactivity will be measured and corrected separately.

Based on the obtained radioactivity level distribution, it is determined whether or not the measurer 1 is contaminated by radioactive substances (step 44). That is, this determination is made by comparing a predetermined value of radioactivity that the person to be measured 1 normally has with the above-mentioned measured value of radioactivity. If the latter value is less than or equal to the former value, the person is not contaminated with radioactive substances, and the test ends there (step 53).

If the latter value exceeds the former value, the person is contaminated with radioactive substances, and the following tests will be conducted on that person. Based on the characteristics of the radioactivity level distribution shown in FIG. 5 and the output signal of the position detector 6 detected during radioactivity measurement, the contamination position in the height direction of the person to be measured 1 +Y)f! ::Determine (step 45).

All the radiation detectors 3 are moved to the determined contaminated position (Y).

Next, the rotating arm 100 is rotated so that the position detector 10
Based on the output signal of step 4, the contamination position (θ) on the circumference with the body axis as the rotation axis is determined in a plane orthogonal to the body axis of the subject 1 (step 45A). Determined contamination position (θ)
At the same time, the radiation detector 3 is moved and the connector 39 is connected to the contact point 35 (step 46B). At this time, connector 3
4 and contact 36 are separated. Radioactivity measurement at the contaminated position fY, θ) is started (step 47). The output signal of the radiation detector 3 is sent to a pulse height analyzer 13 via an amplifier 12. The pulse height analyzer 13 performs pulse height analysis of the input signal and outputs a pulse height distribution of the γ-ray energy spectrum. This γ
The wave height distribution of the line energy spectrum is imported into the central processing unit 16 from the data import interface 15,
It is stored in disk 22 (or magnetic tape 20) (step 48). The person to be measured 1 gets off the bed 38 when the measurement by the radiation detector 3 is completed. Thereafter, the processing described below is performed by the calculation unit 14. The pulse height distribution of the gamma ray energy spectrum necessarily includes a pair of photoelectric peak components P and Compton components C corresponding to one radionuclide. When multiple radionuclides are present, the photoelectric peak component P and the Compton component Cii:
As many radionuclides as exist exist. The lightning peak component P is the photoelectric effect! i! : This is a quantity that represents the size of the peak. The Compton component C is a quantity representative of a pulse height region corresponding to a γ-ray whose energy and Luki are attenuated by co/Noton scattering.

Next, in step 49, the radionuclide causing radioactive contamination is determined. The process of radionuclide determination was carried out in the seventh
This will be explained in detail based on the diagram.

The manually generated wave height distribution of the γ-ray energy spectrum is extracted from the disk 22, and the number of photoelectric peaks is determined (step 49A). The RAM 19 stores values of photoelectric peak energy E for radionuclides as shown in FIG. The photoelectric peak energy E is determined depending on the radionuclide. The photoelectric peak energy Eo is calculated from the wave height distribution (measured value) of the r-ray energy vector taken out from the disk 22 (step 49B). Photoelectric peak energy E
. ? 'i is the energy corresponding to the maximum count value in the photoelectric peak component P. One value of the photoelectric peak energy E stored in the AM 19 is taken out (step 49C).

Photoelectric peak energy E.

and one photoelectric peak energy E (step 4
9D). If Eo = E'r6, is it a pollution source? 6 radionuclides are determined (step 49E). That is, the radionuclide corresponding to the photoelectric peak energy E is the radionuclide that can be claimed. If EO and E are different, RAM19
Another value of photoelectric peak energy E is taken from , and the process of step 49C is performed again. This operation is E,=E
It is repeated until . After one radionuclide has been determined, it is determined whether there is a photoelectric peak for which no other radionuclide has been determined (step 49F).

If the photoelectric peak exists, steps 49A to 4
The process of 9F is repeated (however, 60c.

If there is a radionuclide that has multiple photoelectric peaks, such as in the case of a radionuclide, the number of repetitions will be reduced accordingly. ) These processes are repeated for the number of photoelectric peaks, and a number of radionuclides equal to the number of photoelectric peaks are selected.

The determined radionuclide name is stored on the disk 22.

After all radionuclides that are the source of contamination have been determined, a spectral index calculation is performed (step 50). The calculation of the spectral index will be explained with reference to FIG. disk 22
The peak value of the γ-ray energy spectrum and the name of the radionuclide stored in are input into the central processing unit [16 (step 50A).

The photoelectric peak component P is calculated from the γ-ray energy spectrum for one radionuclide (step 50B), and the Compton component C is added (step 50C). Next, a spectral index S is calculated (step 50D). Spectral index S
is obtained by the operation n of S=C/P. If there are multiple different radionuclides as contamination sources, step 5
The processes 0A to 50D are repeated for the number of radionuclides, and each spectrum index S is calculated. That is, after the calculation in step 50D is completed, it is determined whether there is a radionuclide for which the spectrum index S is not included (step 50E).
). If the radionuclide is present, step 50A~
Repeat step fM of 50D. When it is determined in step 50E that there is no radionuclide that does not contain the spectral index S, the process in step 50 is ended.

The calculation of the spectrum index S described above will be explained in more detail based on a specific γ-ray energy spectrum shown in FIG. FIG. 9 shows NaI (Tt
) This is an example of the pulse height distribution of the γ-ray energy spectrum when a radiation detector is used. Peak E. is the photoelectric peak energy.

The photoelectric peak component P is expressed as the area above a straight line zm connecting four points on the low-energy side of the photoelectric peak and m points on the high-energy side of the photoelectric peak. That is, point l
, m and n'l are the area of the diagonal line quadrupled.

Compton scattering occurs inside the human body of the person to be measured through which the γ-rays pass, and inside the radiation detector 3 through which the γ-rays pass roughly (for example, Na
1 (Tt) in the NaII TA) crystal of the radiation detector). In FIG. 9, Compton scattering that occurs within the human body occurs in a portion where the pulse height is equal to or higher than point h. However, although it is possible to determine the Compton component C at 5 or more points, as will be described later, in the section below the Compton edge energy Ec, the γ-ray B1-value generated by Compton scattering within the radiation detector 3 is It is desirable to use a larger area than the Coppton end (point d), since the accuracy will decrease. When the point is completely exceeded, the light 'L peak component 1)
Accuracy decreases due to the influence of

Therefore, when the Compton component C is determined in the region of the pulse height between points d and 4, the accuracy with respect to the surface of the emitted gamma rays incident on the radiation detector becomes poor. This leads to a reduction in counting time. In this example, as shown in FIG. 9, the area of the γ-ray energy spectrum of the pulse height width between point d and point e, which collapses the pulse height value corresponding to an energy greater than point d, that is, d, The area of the shaded area surrounded by the lines connecting the points e, f, and g is used as compt/component C. The energy Ec of the kombuto/edge corresponding to point d is・
There is a relationship between the photoelectric peak energy Eo and the following formula ("Radiation", Ogawa Roshio).

(Published by Corona Publishing, 1964).

The energy resolution of a NaI(Tt) radiation detector is, for example, 3 inches/inch in diameter and 3 inches thick.
) In the case of crystals, it is about 9chi. Therefore, the pulse when calculating the Compton component C generated in the body of the subject based on the γ-ray energy spectrum obtained from the output signal of the NaI(Tt) radiation detector having NaI(Tt) crystal of this size. The wave height region is preferably between the pulse height values 6 and point e, which correspond to the values of energies E1 and B2 given by the following two equations.

By setting the Copton component C in the region of the pulse height value on the higher energy side than the Compton end, the above-mentioned effect can be obtained. That is, it is not affected by noise components of the radiation detector 3 as in the low energy region. Furthermore, the count value of the pulse height value corresponding to the energy Ec, l:t)R+ is that the γ-ray is N
It does not contain the γ-ray component whose energy is attenuated by Copton scattering when it passes through the aI (T7) crystal, but it contains the γ-ray component whose energy is attenuated by Kobuttono scattering when it passes through the human body. Therefore, when calculating the combdoor component C based on the count value in the above-mentioned pulse peak value region, the resulting spectral index S is the same as the spectral index S that is divided by the combdoor component C in the other region. Compared to this, the measurement accuracy is improved for changes in the depth of the contaminated site.

The reason for this will be explained below. A single energy E as shown in FIG. The pulse height distribution measured in Km when a γ-ray of
If we consider the case where there is no target fluctuation, the result will be as shown in Figure 11. The pulse height value 81 when a single-energy PJO is completely absorbed by the photoelectric effect and the Compton edge energy Ec caused by Copton scattering in the NaI (Tt) crystal.
A pulse height region 82 corresponding to lower energy appears. On the other hand, single energy E. When the γ-rays pass through the human body, they have a continuous energy distribution as shown in FIG. When γ-rays having the distribution shown in Figure 12 are incident on a NaI (Tl) radiation detector, the pulse height distribution ignoring statistical fluctuations is as shown in Figure 13 (which shows the general tendency of the characteristics in Figure 9). )become that way.

The count value of the pulse height below the Copton edge energy Ec is the component 83 due to Compton scattering in the NaI ('L'7) crystal as shown in Fig. 11, and the Copton scattered radiation in the human body is the NaI (T7) crystal. It is the sum of the component 84 based on the pulse that is totally absorbed or generated by scattering within the pulse. In addition, l of the pulse height corresponding to the region from the edge energy EC to the photoelectric peak energy Eo is
The numerical value of 85 is caused by the total absorption of the energy of gamma rays whose energy has been attenuated within the human body. Therefore, in order to accurately know the depth of the contaminated part of the person being inspected, it is necessary to use a pulse in a region equal to or higher than the conogton end energy Ec, which includes only the compensation value that is generated only when γ rays pass through the human body. It is desirable to calculate the Compton component C using the wave height count. If core/component C is determined using the count value of the pulse height in the region below the Compton edge energy Ec, the accuracy of determining the depth of the contaminated area is low due to the influence of Compton scattering that occurs within the radiation detector 3. Become. The above is one method for determining the depth of the contaminated area. When the radioactive nuclide that is the contaminated part is Ichiyu, the spectral index S that determines the contaminated part can be expressed as p- using a component in a lower energy region than the Compton edge energy Ec. Regarding the measurement system to be used, the relationship between the pulse height value of the output of the detector 3 and the component ΔC due to the scattering of Konobuto/component C in the human body, which is used to calculate the spectrum index S, should be determined in advance through experiments, and the meaning of ΔC should be determined in advance. The entire spectrum index S is calculated from the increasing pulse height region. Figure 14 shows the pulse height value and ΔC of the Compton component.
The results of experiments on the relationship between the components are shown. In the energy region lower than the Compton edge energy Ec, the spectral index S
is determined by the ratio of the photoelectric peak component P and the ΔC component of the convex/component C (S - ΔC/p)i.If the spectral index is calculated in the large pulse height region of the connobute/component ΔC, the contaminated area can be determined. Depth can be determined with good sensitivity. In FIG. 14, the largest pulse height value of this combination component ΔC is shown as ΔCmaX. Figure 14 shows the result obtained by subtracting the γ-ray energy spectrum when simulating body surface contamination from the γ-ray energy spectrum when simulating internal contamination, and calculating only the component ΔC due to γ-rays scattered within the human body. be.

The spectrum index S is the area of the shaded area surrounded by the lines connecting points d, e, f, and g in FIG. 9 (component component C) 6 or ΔC in FIG. It is determined by dividing point 2m by the area (photoelectric peak component P) of the shaded area quadrupled by the lines connecting the points n and a.

After completing the calculation of the spectral index S in step 50 of FIG. 3, the reference spectral indexes S1 and S2 are set (
Step 51). The calculation of the reference spectrum indices S1 and S2 will be explained in detail with reference to FIG. 15. First, the thickness of the person to be measured l is read from the disk 22 into the central processing unit 16, and the radionuclide name determined in step 49DK is input (step 51A). The thickness of the subject 1 (thickness of the body 1) is measured separately and then stored in the disk 22 via 8 operation panels.The criterion for determining the depth of the contaminated position, i.e. Quarry of the boundary between the body surface area and the internal body area of Person 1 (hereinafter referred to as area boundary depth) D+
and D2 are determined from FIG. 16 in correspondence with the thickness of the subject 1 described above (step 51B). Region 87 is an intracorporeal region, and regions 86A and 86B are body surface regions.

Body surface area 86A is the back and waist side, body surface area 86
B is the chest and ventral side. Straight line 88A! 5- and 8
8B represents body surface areas 86A and 86B and internal body area 87.
The depth of each boundary is shown.

The body surface regions 86A and 86B are areas such as skin and subcutaneous fat where radioactive substances are difficult to accumulate. The internal body region 87 is a portion where internal organs such as respiratory system organs such as the lungs and digestive system organs such as the stomach, intestines, and liver are located. Radioactive substances tend to accumulate in such areas.

Internal contamination generally occurs by inhaling or swallowing dust containing radioactive materials. Therefore, radioactive substances exist in the above-mentioned respiratory system organs and digestive system organs. These organs are separated by a certain distance from the body surface (the upper surface of the skin). This distance varies depending on the physique, percentage and the thickness of the human body in the direction of radiation measurement. In a standard human body, the height when supine is approximately 20 cm, and the distance from the above organs to the body surface is approximately 3 cm (
Edward M, Smlth et al iJ
, NuclearMedicine, Suppleme
nt A3. Voz, 10+1969, August
st). From the above, when the thickness of the human body is 20 tM, the body surface area 86A is in the range of θ to 3 crrr, which is 3 cm from the front body surface toward the inside of the body, and the body surface area 86B is from the back body surface to the inside of the body. Proceeded 3m in the direction1
It ranges from 7 to 20 crn.

When the radioactive substance is present in the body surface regions 86A and 86B, it is called body surface contamination, and when it is present in the body region 87, it is called internal contamination. The above explanation is based on the physique of a standard human body, but for non-standard human bodies, the ratio of the distance between organs and body surface to the thickness of the human body is similar to that of a standard human body. The boundary depth between the body surface area and the body area is determined as being equal to the ratio of .

When the boundary depth and D2 are determined, in step 51C+FIG. 15), the reference spectrum index S and S2 are determined for each radionuclide based on the characteristics shown in FIG. 17. The characteristics shown in FIG. 17 show the relationship between the spectrum index value determined in advance by the calculation formula S=P/C and the domain boundary method formula.

This %1 life should be determined in advance. The characteristic curve 89 is +
87CB emits 0,662 MeV gamma rays, characteristic curve 90 is L17 MeV gamma rays emitted by 60Co, characteristic curve 91 is 1,33 MeVOr rays emitted by 60Co, and characteristic curve 92 is 0,364 MeV emitted by 131 Engineering.
These are the characteristics corresponding to the respective gamma rays. Since this characteristic differs depending on the radiation detector used and the energy range in which the Combton component is placed, it is necessary to determine this characteristic for each device of the present invention. The relationship between the spectral index and the area boundary quarry shown in FIG.
t) For NaI (Tt) radiation detectors using crystals.For other radionuclides such as 54Mn,
The characteristics shown in FIG. 17 can be easily obtained.

The reference spectrum indices S1 and S2 in step 51CK are the region boundary depth I) determined in step 51B.
, j, i- and D2.

First, a characteristic curve corresponding to the radionuclide present in the contamination source of the person to be measured (determined in step 49D) is selected from FIG. Based on the selected characteristic curve, the spectral indices S and S2 corresponding to each region boundary depth, j, - and D2 can be determined. These spectral indices become reference spectral indices. For example, if the thickness of the person to be measured is 20 on and the radionuclide of the contamination source is 137CB, the reference spectrum index S1 (area boundary depth, = 3C
rn) is the value of point 93A in FIG. 17, and the reference spectrum index S2 (region boundary method D2 = 17crn) is the
This is the value of point 93B in Figure 7. Step 51C like this
When the processing of 8 is completed, the reference spectrum index S and 8
2. Determine the presence or absence of unclaimed radionuclides (step 51D). If the radionuclide is present, repeat steps 51B and 51C. Step 51D
When it is determined that the corresponding radioactive nucleus does not exist, the process of step 51 is completed, and the contaminated area is then determined (step 52 in FIG. 3).

The contents of step 52 are shown in FIG. The spectral index S determined in step 50D for one radionuclide, the reference spectral index S determined in step 51C, and S
2 (step 52A). When the condition of formula (4) shown below is satisfied, it is determined that the radionuclide has internal contamination 52B and S+<S<82...(4), and when the condition of formula (4) is not satisfied. is determined to be body surface contamination 52C. In the next step 52D, the presence or absence of radionuclides that have not been determined is determined. If there is a radionuclide whose contaminated area has not been determined, the determination in step 52A is repeated for that nuclide.

When there are no more radionuclides that have not been identified as contaminated areas,
Taking into account the determination results of step 52A for each radionuclide, the type of contamination of the person to be measured 1 being examined is determined (step 52E). That is, in step 52E,
Internal contamination (first form) considering the determination in step 52A
, determine internal and body surface contamination (second form) and body surface contamination (third form).

When the condition of equation (4) is satisfied for all radionuclides contaminating the person to be measured, the first form of internal contamination occurs. For all radionuclides, if the condition of formula (4) is not satisfied, the third form of body surface contamination occurs.For some radionuclides, the condition of formula (4) is satisfied, If the second condition is not satisfied for other radionuclides, it is the second form of internal body and body surface contamination.When the determination in step 52.E is completed, the determination in step 52 is completed and The identification process for body surface contamination and internal contamination shown in Figure 3 is completed (step 53).Then, the identification results are displayed on the display 31, and the light/internal contamination is identified.
Printed at 30 mm.

When it is determined that the second type of internal and body surface contamination or the third type of body surface contamination occurs, the subject 1 takes a full shower in a shower room (not shown) and cleans the body surface.

This washes away radionuclides attached to the body surface. In the case of body surface contamination, radionuclides are attached to the body surface of the person to be measured, so they can be easily removed by the above-mentioned treatment. A decontamination agent may be applied to the skin instead of soy sauce.

After performing the decontamination treatment, in the case of the third form, step 40
Repeat steps 4-4 to confirm that the radionuclides have been completely removed from the body surface. In the case of the second embodiment, the processes of steps 40 to 5-2 are carried out again to confirm that the body surface contamination has been completely treated.

In the case of the second form, after the above-mentioned re-examination is completed, and in the case of the first form, after the above-mentioned first examination is completed, an examination is performed to determine the internal exposure dose.

The above is a method when contamination with the same radionuclide occurs either internally or on the body surface. Next, a method for determining a contaminated area when body surface contamination and internal body contamination occur simultaneously with a single radionuclide will be described. Internal and body surface contamination with a single radionuclide (secondary
If such a phenomenon occurs, the respective spectral indices S are combined, and the corresponding virtual position is determined to be a contaminated site. Figure 19 shows a state in which multiple contamination of the body and the body surface by a single radionuclide occurs in a cross section perpendicular to the body axis of the subject, and a specific method of area determination will be explained. Subject 1 has internal contamination Y1 and body surface contamination Y. has occurred, and from the measurement results of the radiation detector 3 in the X direction, the virtual depth of the contaminated portion position ta can be recognized.

This is because the spectral index S is determined from the composite gamma ray spectrum of the internal contamination Y1 and the body surface contamination Yo.

As can be seen from FIG. 17, the spectral index S of the body surface contamination component (small area boundary depth) on the side of the radiation detector 3 is small, and the spectral index S of the internal contamination component is thicker.

The area boundary depth of the synthesized spectrum index S is approximately 1 point A from the center of both nails. That is, internal contamination Y1 and body surface contamination Y. ta, which is determined by the strength ratio of , is measured as the depth of the contaminated site. Hereinafter, a method for determining whether the contamination at depth la is due to a combination of body surface contamination and internal contamination will be explained.

To the radiation detector 3': Rotate the rotary arm 1oor (shown in FIG. 1) by Δθ from the contamination position (Y, θ). Fig. 19 shows a state where the radiation detector 3 is movable in the X direction with the radiation detector 3 rotated by 90'' as Δθ.The depth of the contaminated site determined by measurement after rotating the radiation detector 3 is shown. lc.If the contamination A determined before rotating the radiation detector 3 is not complex contamination, Lc corresponds to the depth 7d when the entire point A is viewed from the radiation detector 3 after Δθ rotation.

However, in the case of combined contamination of Yo and Yl, the body surface contamination Yo (in the example of FIG. 19, the radiation detector 3 can be identified as body surface contamination even after Δθ rotation) and the internal contamination Y1 at the depth tb. From the composite spectrum, the contaminated site is determined as point 0 in FIG. 19. That is, when 1c=td, the first form Lc2<td is the $2 form; when tc=ta=0, the third form is the form.

From the above, in the case of contamination by a single radionuclide, internal contamination and combined contamination of internal contamination and body surface contamination can be easily identified from the measurement results at two locations with different measurement angles of the radiation detector 3. .

In this method, the angle at which the radiation detector 3 is rotated may be arbitrary in principle. However, practically, it is desirable to set the angle at 90° or less so that the radiation detector 3 after the S shift can see the body surface contamination position. Although we have explained the method of rotating and moving the radiation detector 3, there is also a method of arranging a plurality of radiation detectors 3 on a coaxial circumference and collating the position depth of the contaminated part based on the results of simultaneous measurement. can also be easily inferred.

An outline of how to estimate internal radiation exposure is explained below. Deposit @leaf equivalent H, which is received by the target organ (for example, another organ) T by the radiation gauze i emitted by the radioactive nuclide j, which is a radiation source, and is present in a certain organ (for example, organ) S in the body. 2
Ding (T'-8) 1. j is the 50-year IF in organ S
It is calculated as the product of the linear variable of IJ and the amount of radiation that organ T receives in one decay stroke. This fL is expressed by the following formula.

Month,. , T('1"←SJ+, 1 = 1.6×1O-10×U,×5EE(-I co--s
)l,j...(5) Here, U is 50 minutes since the radionuclide present in the source organ S has been ingested J[li.
This is the number of times it decays per year. SEE['ll''←S) is the energy absorbed by the target organ 1 when one decay of the radionuclide present in the source organ S is expressed as the radiation quality coefficient. This is the corrected value (N e V / g). 1.6 X 1
0-'. is a constant that converts hieV/g to J/kg.

If multiple radionuclides exist in the source organ S, H
! lo, tI'1'4-8) 1.4 is added for j. Furthermore, when the target organ T is irradiated from a plurality of radiation source organs S, the values are added for each radiation source organ S.

The value of 5EE (T+-8) for each organ is F3nyd
Assuming standard M content and atomic composition of human organs, er et al. performed large-scale monotechnical nomenclature on MIRD Phantom, which mathematically expressed the shape and three-dimensional arrangement of each organ. (W, S, 5nyder.

e(al,;A tabulation of dos
e equivalentp, er m1crocur
ie-day for 5source and targ
et org2ns of an adult for
various radionucliaes
L-5000; 1974).

From the above, in order to determine the internal radiation exposure for each radiation dose, it is sufficient to evaluate the decay rate U of radionuclides within each organ. International Commission on Radiological Protection
on Radiological Protection
In the internal exposure estimation method (IC几P method) proposed by the International Radiological Society (IC几P), the internal transfer of radionuclides from ingestion to excretion is analyzed using a metabolic model, and the The line ring variables are set ( lcH, P Pu1) li
cation 30 Intake limit of radionuclides by workers T
BI part 1; Pergamon Pres
s ; Oxford ; 1980). This metabolic model divides the organs or tissues involved in human metabolic processes into parts (compartments λ) with physiologically identical functions, and shows the internal behavior of radioactive materials as movement between compartments. This is not limited to physically distinct organs such as the stomach and small intestine, but there are also organs that are subdivided into multiple conoparts with different rates of excretion, such as the lungs and liver.Blood is also considered as a compartment.

A metabolic model that takes such compartmentalization into account is
This is established based on the following two assumptions.

(1) Conover) When radionuclides flow into Menoto, they are uniformly mixed and distributed at that moment.

(2) The rate of transfer of radionuclides between compartments is proportional to the total amount of radionuclides present (excluding alkaline earth metals (Ca, Sr, etc.). Excretion is not expressed as an exponential function, but as a power function of time. Furthermore, it is assumed that there are no changes in the transfer rate due to interactions between different radionuclides or morphological changes during transfer into the body.

An example of a metabolic model in the case of inhalation of dust containing BoCo is shown in FIG. 20. This model can be broadly divided into a respiratory system model that runs from the nose through the bronchi to the alveoli, a digestive system model that runs from the stomach to the small intestine and the large intestine, and a circulatory system model that shows the process of excretion via the blood. It is established from The numbers in 0 next to each organ in Figure 20 indicate the number of compartments that serve the same organ.

In the case of inhalation, the rate of deposition in each tissue of the respiratory system changes depending on the particle size distribution of the inhaled dust.

Furthermore, the rate of migration from each tissue depends on the chemical form of the radionuclide. Deposited dust passes into the digestive system by mucus transport and ciliary movement on the surface of the respiratory airways, or dissolves into the blood through the cell walls of the tissues in which it is deposited. Of what passes into the digestive system, a portion is absorbed in the small intestine and transferred to the blood, and the rest is excreted from the body via the large intestine. The rate of absorption from the small intestine is determined by the nuclide and its chemical form. After passing through these processes and migrating into the blood, it follows a fixed route depending on the nuclide, regardless of its chemical form. Since it is estimated from animal experiments that 60Co is concentrated in the liver, only the liver is distinguished and treated as an independent compartment in the excretion route from the blood. The other organs shown in FIG. 20 are the organs of the circulatory system other than the liver. Note that some of the radionuclides that have migrated into the blood can be excreted through a process in which the time delay can be ignored.

The input information necessary for the metabolic model described above is (1) the radionuclide, (2) the amount of radionuclide ingested, (3) the particle size of the inhaled dust, and (4) the radioactivity contained in the inhaled dust. It is a chemical form of a nuclide.

Assuming that the amount of radionuclides present in a certain compartment n is q, (S [B,]), the following differential equation is established from the above assumptions.

1-q, (S=-(λyu+λR)Qyu(j)+In
(Re...(6)) where λ is the conversion member) n is the biologically eliminated rate constant (sec-''), λR is the physical decay constant [,-'), In(riha) (unit time)
The amount of radionuclides flowing into n (B, sec force and L
The time [pond] comes out. Figure 21 shows an example of the behavior of Co in the body, which the inventors found by solving equation (6).
B, when inhaled and ingested.

The method of estimating the internal exposure dose in this example using the metabolic model of the ICRT method described above will be explained below based on FIG. 4.

The subject 1 who has been determined to have been internally exposed through radiation measurement based on the process shown in FIG. 3 lies supine on the bed 38 again. Thereafter, the operator presses the second start button for radioactivity measurement on the operation panel 8.

This command is transmitted to the central processing unit 16 via the panel interface 24. The central processing unit 16 has 1 (
, 0M18 is called, and the internal exposure dose of the person to be measured I is determined according to the program. As mentioned above, the amount of information is (1
) radionuclides, (2) intake of radionuclides, (3) particle size of dust, and (4) chemical form of radionuclides. It is determined by measuring the total amount of residual radioactivity.The intake of radionuclides it5!fl is determined by extrapolating the amount of residual radioactivity at the time of measurement based on the change over time in the amount of residual radioactivity in the whole body obtained from a metabolic model. .Elapsed time from intake to measurement 11
J] can be determined by knowing the distribution status in the body, depending on the other parameters indicating the internal absorption. The particle size of the dust and the chemical form of the radionuclide make it impossible to determine exactly when Victim 1 inhaled the radionuclide.
It is difficult to measure directly. However, these values can be determined as follows. That is, the particle size of dust can be determined according to the aerodynamic radioactivity median diameter (Kanji Takahashi; "Basic Aerosol Engineering") according to the ICRP recommendations when no measured value is available.

p129-132.1972) and 1 μm. The chemical form of the radionuclide is estimated based on the work of the person to be measured 1. For example, if Subject 1 is engaged in periodic inspection of the containment vessel of a boiling water nuclear power plant, there is a high probability that the radionuclides inhaled by Subject 1 are mainly corrosion products of structural materials. expensive. The chemical form of the radionuclide can be assumed to be an oxide because it is an oxide of corrosion products in the structural phase. If the chemical form of the radionuclide is other than an oxide,
This results in an underestimate of the rate at which radioactive nuclides move into the body.

However, the amount of internal exposure will be overestimated and will be on the safe side.

The calculation of the internal exposure amount of the person to be measured l will be specifically described based on FIG. 4. At the same time, the operator presses the second start button and inputs the chemical form 2 of the radionuclide inhaled by the person to be measured and the entire particle size of the dust to the central processing unit 16 from the control panel 8 (step 54). .These information are stored on the disk 22.Then, the connector 34 connects the contact 3
6, the connectors 39 are respectively connected to the contacts 35 (step 55). Next, the central processing unit 16 drives the drive device ff.
: 5 Send the VC command and the radiation detector 3 detects the person being measured.
The moving cart 37 is moved so that it is located at the tip of the head of the person (step 56). When the position setting of the radiation detector 3 is completed, the moving trolley 37 is slowly moved in the direction of the feet along the kite rail 32, and the radiation of the person 1 is measured by the radiation@detector 3. (Step 57). At the same time, the radioactivity measurement position is detected by the position detector 6, and this position signal is stored in the disk 22 via the operation panel 8. The radioactivity signal that is the output of the radiation detector 3 is taken in as it is to the central processing unit 16 via the connector 34,
It can be stored on the disk 22. At the same time, the radioactive signal
After being converted into a wave height value of the gamma ray energy spectrum by the wave height analyzer 13, it is taken into the central processing unit @16. This peak value is also stored on the disk 22 (step 58).

When the radiation detector 3 reaches the tip of the foot of the person to be measured 1, the drive device 5 is stopped and the measurement of radioactivity is completed (step 59). The central processing unit 16 retrieves the peak value of the γ-ray energy spectrum stored in the disk 22 and determines the radionuclide responsible for the internal exposure (step 60). This determination of radionuclides is performed in the same manner as steps 49A to 49F shown in FIG.

60C in the determination process of body surface contamination and internal contamination shown in Figure 3.
It was determined that internal contamination was due to O, and σ et al.
Suppose that it was determined that the only radionuclide causing internal contamination in the experiment was 60CO. Taking as an example the calculation of the internal exposure dose based on the internal contamination of 60CO, the method of calculating the internal exposure dose in this example will be specifically explained.

Based on the radioactivity signal stored in the disk 22, the residual radioactivity t("Co
(Step 61) In order to calculate the intake amount of radionuclide (60Co) by subject 1 based on this residual radioactivity, first, from the time of radionuclide intake to the time of radioactivity measurement. It is necessary to calculate the elapsed time (depending on the work of the person to be measured, the elapsed time can be determined in advance). The elapsed time after ingestion of radionuclides can be determined by taking into account the distribution of radionuclides within the body.

This distribution condition is obtained based on the output signal of the pulse height analyzer 13 at a fixed time of the whole body radioactivity full11. That is, the photoelectric peak energy E in the whole body. It is possible to determine which radionuclide is present in which position based on the relationship between all observations and the measured radioactivity measurement positions. In step 62, the elapsed time after ingesting the radionuclide is calculated. An example of step 62 will be explained with reference to FIG.

Based on the radioactivity signals measured in step 56 and stored in the disk 22, calculate the radioactivity in each of the chest and abdomen of the subject 1 (step 62A).
). The ratio of the amount of radioactivity in the abdomen to the radioactivity in the thorax (hereinafter referred to as the abdomen/thorax residual radioactivity ratio) is calculated (step 62B3). As shown in Table 1, the ratio of residual radioactivity in the chest and abdomen shows the greatest change over time after 60 CO ingestion. This time course is independent of 60CO intake. Therefore, by calculating the residual radioactivity ratio of the abdomen/thorax in Table 1, 60Co
The elapsed time from the time of ingestion to the time of radioactivity measurement can be easily determined. That is, this elapsed time is obtained by calculating the value corresponding to the residual radioactivity ratio of the abdomen/thorax from the characteristics shown in FIG. 23 (step 62C). The characteristics in FIG. 23 show the relationship between the elapsed time after ingestion of Co and the residual radioactivity ratio in the abdomen/thorax.

``Mn'' also has the same characteristics as ``Co.'' In order to measure the elapsed time after ingestion, the characteristic curve shown in FIG. 23 is automatically selected depending on the radionuclide that is the source of internal contamination of the subject determined in step 60. When there are multiple radionuclides as contamination sources in the body, they are generally ingested at the same time. l! Since it seems to have been destroyed, (abdominal radioactivity amount) / ( for either radionuclide)
Radiant radiation on the chest) may be removed.

The amount of radionuclide ingested is determined based on the elapsed time after ingesting the radionuclide and the amount of residual radioactivity at the time of radioactivity measurement (step 63). Characteristics (hereinafter referred to as residual radiation It is called ability characteristic.

An example is shown in FIG. 21) is searched from the disk 22 (or magnetic tape 20). Various residual radioactivity characteristics are determined in advance using the radionuclide, dust particle size, and chemical form of the radionuclide as parameters, and these are stored in the disk 22 (or magnetic tape 20).

Substantially, it is sufficient to simply record the percentage of residual radioactivity for each radionuclide. This is because if the dust particle size is 1 μm and the chemical form is oxide, the measurement will be on the safe side. The residual radioactivity characteristics may not be stored in the disk 22 as described above, but may be calculated each time based on equation (6) in consideration of the three parameters described above.

In this example, the dust particle size is 1 μm and the chemical form is oxide. The inhaled radionuclide is 60Co as determined in step 60. Therefore, disk 22
The characteristics shown in FIG. 21 are searched from. The residual radioactivity corresponding to the elapsed time (time from the time of radioactivity ingestion to the time of measurement) calculated in step 62 is calculated from the whole body characteristic curve shown in FIG. The ratio between this amount of residual radioactivity (inhalation amount IB, per unit amount) and the actual amount of residual radioactivity in the whole body obtained based on the radioactivity signal measured by the radiation detector 3 is calculated.

By extrapolating this ratio and the whole body characteristic curve shown in Figure 21, the intake amount of 60 CO is determined.

The intake amount and radionuclide of the radionuclide determined as described above, and the input particle size of the dust and the chemical form of the radionuclide are applied to the metabolic model shown in FIG. 20, and the cycle variables U, (step 64). That is, it is obtained by integrating a function (which is analytically divided as a solution to equation (6)) that represents the time change in the residual radioactivity of the organ. Specific effective energy of each organ corresponding to radionuclide 5EE (T←S
Search at (step 65)
E(T 4-8l is stored in the disk 22. Specific effective energy 5EE(T 4-8) is
This is specifically shown in the document ``0RNL-5000J'' by Nyder et al. (for 6°Co, i'196).
-81 is input into equation (5) to calculate the committed dose equivalent H6゜4f'l'←S)I'' for the corresponding organ.

The internal exposure dose of the subject L can be determined by adding the products of the committed dose equivalent 11ao4 (T4-8)B of each organ in the body and the weighting coefficient WT given to those organs for each organ. (Step 66). In other words, the internal exposure dose (called effective dose equivalent) Hi
+ is determined by the following formula.

HE=ΣWT-Heo, T(T+-8)-(7) This takes into consideration the fact that even if the same dose is applied to different organs, the degree of influence will differ. The WT values are as shown in Table 2, with five red stamps indicating the highest doses.

If there are radionuclides of class 2a or higher as a contamination source, the true internal exposure dose can be determined by adding up the internal exposure doses due to each radionuclide. Through the above processing, calculation of the internal exposure dose of the person to be measured 1 is completed. The calculated internal exposure dose is printed by a line glitter 30 and displayed on a display 31.

According to this embodiment, since the determination is made based on the Compton component C and the photoelectric peak component P, when the subject 1 is contaminated with radioactivity, internal contamination and body surface contamination can be discriminated with 8 degrees of accuracy. . Therefore, subsequent processing for the subject can be appropriately performed. In addition, when determining internal contamination and body surface contamination, radioactivity should be reduced to 6
111 > 1-Suda Qpu, so the time to restrain runner 1 is short.
Since it is only necessary to measure the radioactivity (at the radioactivity peak position) and perform a wave height analysis, the radioactivity measurement for identifying internal contamination and body surface contamination can be carried out in an extremely short time. The calculation process for identification is also simple, and a determination can be made in a short time. Therefore, even if there are many radioactively contaminated beaches, they can be identified in a short time.

After distinguishing between internal contamination and body surface contamination, in the case of internal contamination, radioactivity measurements are conducted to determine internal exposure, which eliminates the need to detain people with body surface contamination for long periods of time for testing. , the time required to test all contaminated persons will also be reduced. This is because in order to estimate the internal exposure dose, if the elapsed time after ingesting radioactivity is unknown, it is necessary to understand the distribution of radionuclides in the body, and the measured whole body radioactivity signal is required. Wave height analysis must be carried out for In particular, when the amount of radioactivity present in the subject's body is small, wave height analysis of whole body radioactivity signals is
It takes a long time. Therefore, when distinguishing between internal contamination and body surface contamination, it is important to carry out a full wave height analysis of the radioactivity signal of the whole body to determine the internal exposure dose. This will also be done for. In this embodiment, such problems can be solved and radioactivity measurement can be carried out more efficiently. Furthermore, according to this embodiment, it is possible to clearly identify the contamination of a subject who has received radioactive contamination from both internal body contamination and body surface contamination. In such cases, the radioactive nuclei that cause body surface contamination [- After decontamination with a shower, etc.
Since it is possible to carry out radioactivity measurements to determine the internal exposure dose, it is possible to accurately increase the internal exposure dose while eliminating the effects of radioactivity from body surface contamination.

Based on the ratio of residual radioactivity in the abdomen/thorax, the elapsed time after ingestion of the radionuclide can be easily and accurately determined. Therefore, the radioactive nucleus; I'! The entire intake of li can be easily rewarded and dispensed.

〔Effect of the invention〕

According to the present invention, even if body surface contamination and internal contamination occur simultaneously, if the same radionuclide is used, body surface IFU can be easily detected.
It is possible to determine whether there is contamination or not, and to quickly take appropriate measures for the person being measured.

[Brief explanation of drawings]

Figure 1 shows radioactivity '611 which is a preferred embodiment of the present invention.
Figure 2 is a system diagram of the computer system shown in Figure 1, Figure 3 is a fraudulent system diagram of the calculation unit shown in Figure 1. Handling +
111-Lk, FIG. 4 is a flowchart showing the procedure for internal exposure dose calculation carried out by the central processing unit shown in FIG. 1, and FIG. 5 is the radioactivity distribution in the height direction of the subject. FIG. 6 is an explanatory diagram showing the photoelectric peak energy of the radionuclide; FIG. 7 is a flowchart showing the procedure for determining the radionuclide shown in FIG. 3;
FIG. 8 is a flowchart showing the calculation procedure of the spectrum index S shown in FIG. Figure 1O is a characteristic diagram showing the energy of γ rays input to the radiation detector, and Figure 11 is a characteristic diagram showing the energy of γ rays input to the radiation detector.
Figure 12 is a characteristic diagram showing the pulse height distribution output from the radiation detector when a radiation is input, and Figure 12 is a characteristic diagram showing the energy distribution of the γ-rays that passed through the human body when the human body was irradiated with the γ-rays shown in Figure 10. Fig. 13 is a characteristic diagram showing the Norls wave height distribution that is output from the radiation detector when γ@ shown in Fig. 12 is input, and Fig. 14 is a characteristic diagram showing the Norls wave height distribution of the Figure 15 is a flowchart of the reference spectrum index calculation shown in Figure 3. Figure 16 is a characteristic diagram showing the relationship between human body thickness and region boundary depth. , Fig. 17 is a characteristic diagram of the relationship between area boundary depth and spectral index, Fig. 18 is a flowchart of the contaminated area determination in Fig. 3, and Fig. 19 is a diagram of the composite contamination state and radiation detection of the cross section of the subject. FIG. 20 is an explanatory diagram showing an example of a substitute model; FIG. 21 is a characteristic diagram showing the relationship between 1) J] and residual radioactivity in the body after ingesting 60Co; Figure 22 is the fourth
The flowchart for calculating the elapsed time after radionuclide ingestion shown in Figure 23 is the elapsed time after radionuclide ingestion.
The relationship between this and the residual radioactivity ratio in the abdomen/thorax! This is a 14-note figure. l...person to be measured, 2...radioactivity measuring device, 3...
Radiation detector, 4...drive! IIJJ Division, 5...Kaku#
JJ device, 6...position detector, 7...operation unit, 8.
...Operation panel, lO...total 11119b, 13...
- Wavefront analyzer, 14... Arithmetic unit, 16... Central processing unit, 18... J and OM, 19... 1 and AM, 22
...disk, 29...display unit, 31...display, 34.39...connector, 38...bed,
100... Rotating arm, 101... Gear, 102...
... Deceleration motor, 103... Movement position of radiation detector, 104... Rotating house 5 figure foil 6 section fat '1 figure 8 flash letter q figure / - Nunome store W) Figure 10E. Toya Geki Energy One Tank 11 Diagram / Ma 1 Shijin Fu Toshin - 1 I Nao (4 dry sounds 1 Shin λsu Marriage 12 Diagram EO t Latitude Energy N- Taku 13 Diagram Ha 1 Shi) Ryo High Price Condolence 14 -Japanese/Ma) Shino, 1? □L value monk 16 ward 1 10 "Button Yato,. 4 Tsuba C creation I'm figure side revise doo saku cm) Condolence 18 sentry 2i figure 60 co 1?i-temporary sand dust time (dαzo 9 cut 221'
l to ν3 and 1

Claims (1)

[Claims] 1. Measure the radioactivity of the subject, calculate the γ-ray energy spectrum of the measured radioactivity signal, and determine the radioactivity of the subject based on the ratio of the photoelectric effect component and the kelp/scattered component of the γ-ray energy spectrum. In a radioactivity measurement device that identifies body surface contamination and internal body contamination, the contaminated site of the subject is measured simultaneously or independently from two or more different positions, and the contaminated positions obtained from each measurement result are compared with each other. A method for identifying radioactivity in the body, which is characterized by identifying combined contamination of body surface contamination and internal contamination.