JP2002257996A - Neutron generating device - Google Patents

Abstract

PROBLEM TO BE SOLVED: To generate neutrons in such a way that the ratio of thermal neutrons to epithermal neutrons is small, and that the absolute quantity of the epithermal neutrons is large. SOLUTION: The moderator range is surrounded by a reflector comprising a substance having less moderating power than moderator 2, and a neutron source 1, is disposed more on the opposite side to a neutron radiation direction to the center of length of the moderator in the neutron radiation direction in the moderator 2. In a neutron non-destructive inspection device utilizing epithermal neutrons, noise by thermal neutrons at the time of inspection is reduced to improve the reliability of inspection, and the inspection time can be reduced.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

The present invention relates to a neutron generator used for a neutron nondestructive inspection device.

[0002]

2. Description of the Related Art A neutron nondestructive inspection method irradiates a test object with neutrons and measures the attenuation of a neutron flux after transmission due to neutron absorption of a specific element, or the radiation due to a nuclear reaction with a specific element. This is a method for evaluating the presence or absence and amount of the specific element in the inspection object.

[0003] For example, in Japanese Patent Application Laid-Open No. 5-288888,
As shown in FIG. 2, a nuclear fuel pellet 20 coated with zirconium diboride is irradiated with thermal neutrons 22 from a neutron source 21, a neutron capture gamma ray 23 is measured by a gamma ray detector 24, and boron 10 in the nuclear fuel pellet coating is measured. how to measure nondestructively the number density of (less ¹⁰ B) are shown. Here, the neutron capture gamma rays mean not only gamma rays due to the (n, γ) reaction but also all emitted gamma rays due to neutron absorption.

Further, the fact that a non-destructive, not the fuel cladding tube of a nuclear fuel pellet 20 which is an object to be inspected is destroyed by neutron irradiation, even with respect to ¹⁰ B, 10 ¹⁰ boron atoms in test It means that no more than one boron atom is lost. Thermal neutrons are neutrons whose kinetic energy is less than about 1 eV.

Here, the nondestructive inspection method disclosed in Japanese Patent Application Laid-Open No. 5-288888 utilizes the fact that the thermal neutron absorption cross section of ¹⁰ B is very large as shown in FIG.

In the above method, it is necessary to measure ¹⁰ B neutron capture gamma rays with a magnitude that is statistically significant with respect to noise during measurement. Here, the noise at the time of measurement includes neutron capture gamma rays due to constituent elements other than ¹⁰ B of the inspection object, and radiation from structures around the inspection apparatus.

Therefore, in the above method, the test object is irradiated with thermal neutrons having a large neutron absorption cross section of ¹⁰ B and a high probability of emitting neutron capture gamma rays. Other than ¹⁰ B, generally, the smaller the energy of neutrons, the larger the neutron absorption cross section of the substance. Therefore, in the nondestructive inspection method using the neutron absorption reaction, thermal neutrons are used for the above-described reason.

As a method for obtaining thermal neutrons used in such nondestructive inspection, neutrons having a kinetic energy of several MeV or more generated by fission or the like are converted into light water, heavy water, paraffin,
There is known a method of reducing the speed by elastic scattering or the like in a moderator composed of a light element such as graphite.

In Japanese Patent Application Laid-Open No. 2000-56098,
This shows a low-energy neutron irradiation apparatus that can obtain a required number of low-energy neutrons even when a radioisotope having a relatively small output or an electronic neutron generator is used as a neutron source.

For this purpose, the low-energy neutron irradiation device surrounds the neutron source and is made of lead or beryllium.
It comprises a neutron multiplication deceleration region made of a substance having a large (n, 2n) reaction cross-sectional area, and an irradiation container surrounding the neutron multiplication region and having only a neutron irradiation direction open. The irradiation container is made of a neutron absorbing substance such as a hydrogen compound or concrete.

[0011]

SUMMARY OF THE INVENTION The above-mentioned Japanese Patent Application Laid-Open No. 5-288.
As in No. 887, the neutron non-destructive inspection method using thermal neutrons is based on the fact that the element to be measured, which absorbs irradiation neutrons and causes a nuclear reaction (hereinafter referred to as neutron absorption element), has a certain number density in the irradiation direction. It is effective only when it is thinly distributed.

In such a case, part of the irradiated thermal neutrons is absorbed and the rest is transmitted without being absorbed, and the absorption ratio is correlated with the density of the neutron absorbing element. However, if the neutron-absorbing elements are distributed with a certain number density and thickness above a certain level, even if the irradiated thermal neutrons penetrate a little deeper from the surface, they will eventually be all absorbed by the neutron-absorbing elements, and the absorption ratio will be It is not possible to measure the density of the neutron absorbing element without being correlated with the density of the neutron absorbing element.

As described above, when the neutron absorbing element has a certain number density and a certain thickness or more and the thermal neutron irradiation does not provide a correlation between the thermal neutron absorption ratio and the neutron absorbing element density, a higher value is obtained. By irradiating neutrons with kinetic energy, a correlation between the neutron absorption ratio and the neutron absorption element density can be obtained.

This is because the isotope composition ratio of ¹⁰ B is 20% and 5
A sample in which 0% B ₄ C powder is sealed in a stainless steel tube (hereinafter referred to as 20% B ₄ C and 50% B ₄ C, respectively) is irradiated with neutrons, and the difference in the number of ¹⁰ B neutron capture gamma rays emitted is measured. ^A description will be given using calculation results simulating a system for identifying the isotope composition ratio of ¹⁰ B.

The stable isotopes of boron are ¹⁰ B and boron 11
(Hereinafter ¹¹ B) is present, but does not absorb most ¹¹ B neutron, if ^{the 10} B has absorb neutrons, the neutron capture gamma ray is observed by the reaction of the following has been confirmed experimentally.

[0016] ^{10 B + n → 7 Li *} + 4 He 7 Li * → 7 Li + gamma above reaction formula n is neutron, ⁷ Li * lithium excited state, ⁴ the He helium, ⁷ Li basal lithium Indicates a state. Gamma rays are emitted as a result of a state transition from the excited state of lithium to the ground state.

FIG. 4 shows that the energy of the irradiation neutron is 0 eV to 1 e.
The figure shows the difference between the measured numbers of neutron capture gamma rays of 20% B ₄ C and 50% B ₄ C when changed to 0 MeV. As can be seen from FIG. 4, since most of the irradiated neutrons are absorbed by ¹⁰ B in the thermal neutron region, ¹⁰ B
The correlation between the neutron absorption ratio and the density of ¹⁰ B becomes smaller, and therefore the difference in the measured number of neutron capture gamma rays emitted from 20% B ₄ C and 50% B ₄ C becomes smaller.

When the energy of irradiated neutrons is larger than 10 keV, the neutron absorption cross section of ¹⁰ B becomes smaller, and the absolute number of neutron capture gamma rays becomes smaller, so that 20% B ₄ C and 50% B ₄ C The difference in the number of neutron capture gamma rays becomes smaller.

From these facts, in order to discriminate the isotope composition ratio of ¹⁰ B in B ₄ C from the measured number of neutron capture gamma rays, the irradiation neutron energy may be set to 10 to 100 eV.

Since the neutron absorption ratio varies depending on the B ₄ C powder encapsulation conditions such as thickness and density and the composition ratio of ¹⁰ B, the optimum irradiation neutron energy depends on the configuration and shape of the inspection object. , Approximately in the range of 1 eV to 1 keV.

The same method can be applied to the case where the inspection object is other than B ₄ C, that is, when the neutron absorbing element is an element other than ¹⁰ B. Hereinafter, a neutron having a kinetic energy in this range is referred to as an epithermal neutron.

Neutrons obtained from a spontaneously fissile radioisotope or a neutron source utilizing a fusion reaction or a (p, n) reaction have a kinetic energy of several MeV or more,
In the nondestructive inspection, this is used by reducing the energy with a moderator composed of a light element as described above.

Neutrons lose energy mainly by elastic scattering with light elements in the moderator, and the amount of energy lost depends on the number of elastic scatterings in the moderator. The number of this elastic scattering depends on the distance that the neutron travels through the moderator, so by changing the moderator thickness,
The neutron energy after deceleration can be changed to some extent.

However, the energy lost by neutrons in one elastic scattering is not constant but distributed with a certain width.
In addition, since the neutrons change the traveling direction at random by elastic scattering, the distance traveled in the moderator also changes at random, and accordingly, the number of elastic scatterings also changes.

Even if neutrons having a constant energy are incident on the moderator, the energy of the neutrons obtained after deceleration is not constant but has a distribution. Therefore,
When trying to obtain epithermal neutrons, usually thermal neutrons and fast neutrons having higher energy than epithermal neutrons are also obtained.

The problem in the case of using a moderator composed of a light element simply as in the prior art is that the neutrons that have undergone elastic scattering repeatedly become thermal neutrons, and therefore only the thickness of the moderator is changed. In this case, the number of obtained epithermal neutrons is limited, and the ratio of simultaneously obtained thermal neutrons is increased.

In the nondestructive inspection using epithermal neutrons, as shown in FIG. 4, since the thermal neutron absorption ratio does not correlate with the density of the neutron absorbing element, it is irrelevant to the inspection. It is desirable that the amount be as small as possible because it causes noise during measurement.

An object of the present invention is to provide a neutron generator in which the ratio of thermal neutrons to epithermal neutrons is as small as possible and the absolute amount of epithermal neutrons is large.

[0029]

A neutron generator according to the present invention comprises a neutron source, a moderator arranged so as to surround the neutron source, and a reflector surrounding the moderator. The neutron source is made of a substance having a moderation power smaller than that of the moderator, and the neutron source is disposed closer to the opposite side to the irradiation direction than the center of the moderator in the neutron irradiation direction length.

[0030]

FIG. 1 shows a first embodiment of the present invention, in which a cylindrical stainless steel container in which californium 252 (hereinafter referred to as ²⁵² Cf) as a neutron source is disposed. The periphery of the moderator region filled with heavy water is surrounded by a reflector 3 made of nickel.

At this time, one of the cylindrical bottom surfaces of the moderator region is not covered by the reflector 3, and in the nondestructive inspection, the cylindrical bottom surface of the moderator not covered by the reflector 3 (hereinafter referred to as the irradiation surface). The test object is placed in the vicinity and irradiated with neutrons. The neutron source is disposed at the radial center of the moderator cylindrical shape and closer to the bottom surface on the side opposite to the irradiation surface than the moderator longitudinal center c.

Next, the effect of this embodiment will be described. FIG. 5 shows that the cylindrical bottom radius (hereinafter referred to as the moderator radius) of the moderator region is 6 cm,
The cylindrical length of the moderator area (hereinafter moderator length L) is 8 cm,
The neutron flux obtained when the neutron source is arranged at a position 5 cm from the irradiation surface is referred to as a reflector 3 using nickel, heavy water,
This is a case where the case is made of light water and a case where no reflector is used.

It should be noted that 10 eV when a nickel reflector is used
It is expressed as a relative value when 〜100 eV neutron flux is set to 1. FIG.
From, by using the nickel reflector, the thermal neutron ratio of 0 to 1 eV is small, and 10 eV to 100 eV and 100 eV
It can be seen that the epithermal neutron flux of 1 keV is large.

FIG. 6 shows the neutron flux when the moderator length direction position of the neutron source is changed when the moderator length is set to 10 cm. FIG. 6 shows that the epithermal neutron flux is larger when the neutron source is arranged closer to the bottom surface opposite to the irradiation surface than the length center of the moderator.

Next, it will be shown that it is a necessary condition for achieving the object of the present invention to dispose the neutron source closer to the bottom surface opposite to the irradiation surface than the length center of the moderator.
FIG. 7 shows the change in the neutron flux when the position of the neutron source in the moderator length direction is changed when the moderator length is set to 14 cm, in comparison with the case where the moderator length is 10 cm. Things.

FIG. 7 shows the neutron source position relative to each moderator length. The neutron flux has a moderator length of 10 cm and the neutron source position is 10 cm away from the irradiated surface.
The neutron flux is normalized so that the neutron flux is 10 to 100 eV when it is set to cm. As shown in FIG. 7, when the moderator length is increased, the neutron source position where the epithermal neutron flux is maximized changes toward the irradiation surface.

However, when the moderator length is increased, the maximum value of the epithermal neutron flux hardly changes, while the thermal neutron flux obtained at the same time is relatively large. In other words, in order to minimize the ratio of thermal neutrons to epithermal neutrons, the position of the neutron source at which the epithermal neutron flux is maximized is closer to the bottom surface on the side opposite to the irradiation surface than the moderator length center. This indicates that the length of the moderator must be as follows.

The above effects can be explained as follows. Neutrons are generally emitted isotropically from a neutron source, and even if they are emitted in one direction, they may travel in a direction different from the irradiation surface direction by performing elastic scattering in the moderator.

By directing such neutrons toward the irradiation surface by the reflector, the neutron flux on the irradiation surface can be increased. However, as the opposite side region becomes longer, for example, as shown in FIG. 11, the neutrons reflected by the reflector on the opposite side to the irradiation surface and reaching the irradiation surface travel through the moderator, and the direct irradiation from the neutron source is performed. Variations in the distance that the neutrons that reach the irradiated surface travel through the moderator are increased such that the difference in the distance B at which the neutrons that reach the surface travel through the moderator is increased.

Therefore, the energy of the neutrons reaching the irradiation surface varies greatly. This means that when trying to increase epithermal neutrons, thermal neutrons also increase at the same time.

For this reason, as in the present invention, by reducing the area on the opposite side and increasing the number of neutrons that decelerate in the moderator area between the neutron source and the irradiation surface, the neutrons travel in the moderator with a variation in distance. And the variation in neutron energy is suppressed, so that epithermal neutrons can be generated efficiently.

When the reflector has a moderating ability equal to or greater than that of the moderator, the neutrons are decelerated in the reflector and become thermal neutrons. Therefore, the epithermal neutron flux on the irradiation surface must be increased. Can not. Here, the deceleration ability is
It is the product of the average energy lost by neutrons in one elastic scattering and the scattering cross section.

Therefore, it is necessary that the reflector is made of a substance having a moderation power smaller than that of the moderator, a large neutron scattering cross section, and a small neutron absorption cross section. For this reason, as in JP-A-2000-56098, it is difficult to effectively obtain epithermal neutrons when the moderator region is surrounded by an irradiation container made of a neutron absorbing material such as concrete or a hydrogen compound.

Although nickel is used as the reflector in this embodiment, for example, stainless steel, which is an alloy of a transition metal such as nickel, iron, and chromium, and zircaloy, which is a zirconium alloy, can also be used.

By the way, in this embodiment, 10 eV to 10 eV
Although the moderator shape that maximizes the neutron flux of 0 eV has been described as an example, when a higher energy epithermal neutron is required, the shape is adjusted so that the neutron flux of 100 eV to 1 keV is maximized. Decide. In addition, 10eV ~ 100eV or 100eV
The energy division of eV to 1 keV is for convenience, and the optimal energy division is used according to the purpose.

In the above embodiment, heavy water containing a large amount of deuterium was used as the moderator. For example, polyethylene containing a large amount of hydrogen or beryllium may be used. When polyethylene is used as the moderator, when the moderator length is 2 cm, the change in the neutron flux when the position of the neutron source in the moderator length direction is changed, when the moderator length is 3 cm FIG. 8 shows the result of comparison with FIG. FIG. 8 shows the neutron source positions relative to the respective moderator lengths.

The neutron flux has a moderator length of 2 cm,
10 eV to 100 eV when the neutron source position is 2 cm from the irradiated surface
The neutron flux is standardized to be 1. From FIG.
Even if polyethylene is used as the moderator, if the moderator length is increased, the neutron source position where the epithermal neutron flux becomes maximum changes near the irradiation surface, but the maximum value of the epithermal neutron flux does not change much. Only the thermal neutron flux obtained at the same time increases.

Therefore, in order to reduce the thermal neutron ratio and increase the epithermal neutron flux, it is necessary to arrange the neutron source closer to the bottom surface on the side opposite to the irradiation surface than the center of the moderator length. . Similarly, in the case of using beryllium, the comparison between the moderator length of 10 cm and the moderator length of 14 cm is shown in FIG.
Shown in

When polyethylene is used as the moderator, the ratio of thermal neutrons is larger than when heavy water is used as the moderator, but the absolute amount of epithermal neutrons is increased, and There is an advantage that the area can be reduced.

In addition, when beryllium is used, the absolute amount of the obtained epithermal neutron flux is slightly smaller and the ratio of thermal neutrons is slightly larger than when heavy water is used, but it is easy to handle as a material. There is an advantage that there is.

In the above embodiment, the case where a point-like neutron source is used has been described. On the other hand, FIG. 9 shows an embodiment in which a wire-like neutron source is used. The wire-shaped neutron source is parallel to the moderator region cylindrical length direction, and at the cylindrical radial center, the center of the wire-shaped neutron source is closer to the bottom surface on the opposite side of the irradiation surface than the moderator region cylindrical length center. To be placed.

In the above description, the reflector is arranged so as to cover only the moderator region. However, in order to increase the irradiation neutrons on the inspection object, the reflector is decelerated when the inspection object is relatively small. It is also conceivable to arrange the reflector so as to cover the material region and part or all of the inspection object.

FIG. 10 shows an example in which the reflector is arranged so as to surround the side surface of the object to be inspected when the object to be inspected is an elongated rod like a control rod of a nuclear reactor. By arranging the reflectors in this manner, neutrons emitted from the irradiation surface and traveling in a direction different from that of the inspection object can be directed to the inspection object, and the amount of irradiation neutrons can be increased.

As shown in FIG. 10, if necessary, for example, when arranging a radiation detector, a hole is formed in a part of the reflector from the side of the inspection object to the opposite side. Also,
When the object to be inspected is very small, the amount of irradiated neutrons can be further increased by adopting a configuration in which the object to be inspected is arranged in a closed space including the moderator and the reflector.

[0055]

The neutron generator of the present invention generates neutrons so that the ratio of thermal neutrons to epithermal neutrons is small and the absolute amount of epithermal neutrons is large.

Thus, a non-destructive inspection is carried out by irradiating an epithermal neutron to an inspection object composed of a substance in which the neutron absorbing element is distributed so as to have a thickness and a number density such that thermal neutrons cannot pass therethrough. In the case, the inspection time is short, and the noise at the time of inspection by thermal neutrons is reduced,
The reliability of inspection can be improved.

[Brief description of the drawings]

FIG. 1 is a conceptual diagram of an embodiment of the present invention.

FIG. 2 is an explanatory diagram of a neutron nondestructive inspection method.

FIG. 3 is a neutron absorption cross section of boron 10;

FIG. 4 is a graph showing the irradiation neutron energy dependence of the difference in the neutron capture gamma ray emission ratio of B ₄ C depending on the composition ratio of boron 10;

FIG. 5 is a diagram showing an effect of a reflector when heavy water is used as a moderator.

FIG. 6 is a diagram showing a relationship between a neutron source position and a neutron flux when heavy water is used as a moderator.

FIG. 7 shows a neutron source position when heavy water is used as a moderator,
It is a figure which shows the relationship between a moderator length and a neutron flux.

FIG. 8 is a diagram showing a relationship between a neutron source position, a moderator length, and a neutron flux when polyethylene is used as a moderator.

FIG. 9 is an embodiment when a wire-like neutron source is used.

FIG. 10 is a view of an embodiment in which a reflector is arranged at a position distant from the moderator.

FIG. 11 is a diagram illustrating an effect of the present invention.

FIG. 12 is a diagram showing a relationship between a neutron source position, a moderator length, and a neutron flux when beryllium is used as a moderator.

[Explanation of symbols]

DESCRIPTION OF SYMBOLS 1 ... Neutron radiation source, 2 ... Moderator, 3 ... Reflector, 4 ... Wire neutron source, 5 ... Inspection object, 6 ... Radiation detector, 20 ...
Nondestructive inspection object, 21: neutron source, 22: thermal neutron, 23: neutron capture gamma ray, 24: gamma ray detector.

 ──────────────────────────────────────────────────の Continuing on the front page (72) Inventor Hajio Aoyama 7-2-1, Omika-cho, Hitachi City, Ibaraki Prefecture F-term in the Electric Power and Electricity Research Laboratory, Hitachi, Ltd. 2G075 CA38 CA39 DA07 FA06 FA18 FC14 GA15 GA21

Claims (6)

[The claims] 1. A neutron source, a moderator arranged so as to surround the neutron source, and a reflector surrounding the moderator and having an opening at a position in a neutron irradiation direction, the reflector comprising: The neutron source is formed of a substance having a moderation power smaller than that of the moderator, and the neutron source is disposed closer to the side opposite to the neutron irradiation direction than the center of the length of the moderator in the neutron irradiation direction. Neutron generator. 2. The neutron source linearly extends in a neutron irradiation direction, and at a position in the neutron irradiation direction of the neutron source,
2. The neutron generator according to claim 1, wherein a center of the neutron source in a length direction is disposed closer to a side opposite to the neutron irradiation direction than a center of a length of the moderator in a neutron irradiation direction. 3. 3. The neutron generator according to claim 1, wherein the neutron source is a radioisotope that emits neutrons by at least one of spontaneous fission and other nuclear reactions. 4. The neutron generator according to claim 1, wherein the moderator includes a substance containing at least one of hydrogen, deuterium, and beryllium as a constituent element. 5. The neutron generator according to claim 1, wherein the reflector includes a substance containing at least one of transition metals and zirconium as a constituent element. 6. The neutron generator according to claim 1, wherein the reflector is open only in a neutron irradiation direction.