JP2001281384A - Monitoring method of critical approach and measuring jig for monitoring - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a monitoring method of critical approach of neutron multi plication system capable of automatically judging whether a system including nuclear fuel approaches the criticality or not and preventing a criticality acci dent, and a method capable of estimating the generation ratio of neutron and the outline of the change with time, even if the criticality accident has ocurred, from the outside of a building in the accident site. SOLUTION: A neutron detector is arranged on the side surface or inner part of a fuel container. When the relation of a neutron flux (ϕ0) with a large effective magnification to an effective multiplication factor (k0) is ϕ0=a0+b0/(1-k0), and the relation of a neutron flux (ϕ) in a small effective magnification with an effective multiplication factor (k) is ϕ=a+b/(1-k) (wherein a0, b0, a and b are constants), the area where (ϕ0/ϕ) and 1/ρ0 (wherein ρ0=(1-k0/k0) have a substantially linear relation is determined from the relation of (ϕ0/ϕ) with ϕ0, ϕ, k0, and k, ϕ0 and ϕ are measured to monitor the change of k0, and when k0 exceeds a fixed value, an alarm is issued.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method and apparatus for monitoring critical proximity of a neutron multiplication system including a nuclear fuel.

[0002]

2. Description of the Related Art In general, a strict design is made so that a fuel storage pool, a fuel transport container, a spent fuel reprocessing facility, etc. of a nuclear reactor do not become critical, and various methods for confirming subcriticality are proposed. These systems ensure sufficient criticality safety.

On the other hand, in the new fuel processing facility, in order to keep the fuel with the highest enrichment to be handled sufficiently subcritical as an optimal deceleration condition, the shape and size management mainly specifying the shape and size of the fuel storage unit, the enrichment, It is designed to be controlled by mass, concentration, etc.

[0004] The present inventors have included the present inventors regarding a subcriticality measurement method for a neutron multiplication system, particularly for a system of a spent fuel assembly alone or for a system in which a large number of spent fuel assemblies are arranged. Many methods have been proposed by many researchers.

However, there have been few inventions relating to a subcriticality measuring method for a fuel solution system in which the neutron multiplication factor changes with time. Generally, such a system is designed by the above-described methods such as shape / dimension management, volume management, mass management, and density management. Although these designs are critical safety designs with a sufficient margin, it is also important to realize a method for strictly preventing criticality accidents and ensuring higher criticality safety.

In realizing such a method, it is desirable to have not only means for generating an alarm immediately after the occurrence of an accident but also means for determining whether or not a system including nuclear fuel is approaching criticality. It is.

[0007]

SUMMARY OF THE INVENTION The present invention can automatically determine whether a system containing nuclear fuel is approaching criticality as long as a criticality monitoring device is operating, and can prevent a criticality accident from occurring. Criticality monitoring method of neutron multiplication system that can be used.Also, it is possible to predict criticality accidents in advance by sampling the activation measurement jig periodically and grasping the change of neutron multiplication system. Even if a criticality accident occurs, it is a method of estimating the neutron generation rate and the outline of its change over time from outside the building at the accident site. b) The purpose of the present invention is to provide a measuring instrument having a simple configuration and calibration work, (c) easy maintenance and management, and (d) extremely low cost.

[0008]

In order to achieve the above object, a critical proximity monitoring method according to a first aspect of the present invention provides a neutron detector in which a neutron detector is arranged on the side or inside of a fuel container, and the neutron flux when the effective multiplication factor is large. The relationship between (φ ₀ ) and the effective multiplication factor (k ₀ ) is φ ₀ = a ₀ + b ₀ / (1-k ₀ ), and the neutron flux (φ) and the effective multiplication factor (k) when the effective multiplication factor is small. Relationship with φ
= A + b / (1−k) (where a ₀ , b ₀ , a, b are constants), and from the relationship between (φ ₀ / φ) and φ ₀ , φ, k ₀ , k, (φ ₀ / Φ) and (1 / ρ ₀ ) (where ρ ₀ = (1
−k ₀ ) / k ₀ ) are found to be in a substantially linear relationship,
It is characterized in that φ ₀ and φ are measured to monitor the change of k ₀ , and an alarm is issued when k ₀ exceeds a certain value.

Note that the neutron flux ratio φ ₀ / φ is a ₀ , b
₀ , a, b, and R (where R = k ₀ / k) are constants, and 1 /
The following equation specifically shows that there is a substantially linear relationship with ρ ₀ .

(Φ ₀ / φ) = (b ₀ / b) ｛(1-a
/ Φ) / (1-a ₀ / φ ₀ )｝ × [1+ (1-R) / ρ
₀ ] A critical proximity monitoring method according to claim 2, according to the critical proximity monitoring method according to claim 1, wherein in the fuel region arranged in the neutron non-moderator space, the fuel region when the effective multiplication factor is k _0. Is a system in which a neutron reflector is mounted on the side surface of the fuel region, and the fuel region when the effective multiplication factor is k is a system in which the neutron reflector is not mounted on the side surface of the fuel region.

[0011] The critical proximity monitoring method according to claim 3 is based on claim 1.
According to the critical proximity monitoring method, in the fuel region arranged in the neutron moderator space, the fuel region in the case where the effective multiplication factor is k ₀ is a system in which a neutron absorber is not attached to the side surface of the fuel region. When the magnification is k, the fuel region is a system in which a neutron absorber is mounted on the side surface of the fuel region.

According to the present invention, a neutron reflector is attached to and detached from one kind of fuel system having a constant fuel composition, concentration and shape regardless of whether the fuel region is in the air or in the water. By attaching or detaching the neutron absorber, the change in the neutron flux can be expressed as a relative ratio greater than 1, and as described in detail in Example 1, it can be correlated with the reciprocal of the reactivity and the linear relationship. As a result, the execution multiplication factor is obtained, so that the critical proximity situation can be clearly confirmed, and an alarm can be generated accurately and easily as needed.

According to a fourth aspect of the present invention, a critical proximity monitoring method is provided.
In the critical proximity monitoring methods of (1) to (3), a moderator for a neutron detector having a thickness of at least 3 cm in water equivalent is disposed on the neutron reflector mounting / removing side or the neutron absorbing mounting / removing side surrounding the neutron detector. The configuration is characterized in that the influence of the change of the neutron field due to the attachment and detachment of the body is suppressed.

The thickness of the water or hydrogen-containing substance surrounding the neutron detector is preferably at least 3 cm in terms of water, and preferably the thickness up to the vicinity of the detachable neutron reflector or neutron absorber.

According to the present invention, the effect of attachment / detachment of the reflector or absorber is reduced by the neutron deceleration and absorption effects of hydrogen, and the constants (a ₀ , a) and (b ₀ , b) appearing in the basic formulas Can be reduced, and no extreme error occurs in the calculation of these constants obtained by the calculation.

A critical proximity monitoring method according to a fifth aspect of the present invention is the first aspect.
In the critical proximity monitoring methods of (1) to (3), a neutron transport / calculation method for obtaining an effective multiplication factor k and a neutron flux φ for at least two types of fuel concentrations in a measured system in which a neutron source is arranged.
A diffusion calculation is performed to find a place where the absolute values of the ratios (a / b) and (a ₀ / b ₀ ) of the parameters become smaller with respect to a place apart from the neutron source by a distance equal to or more than a certain distance. A neutron detector.

According to the present invention, the factor {(1−a / φ) / (1−a ₀ / φ ₀ )} in the expression relating to the neutron flux ratio described in claim 1 is approximately 1, so that the neutron flux ratio (Φ ₀ / φ) and (1 /
ρ ₀ ) gives a more perfect linearity.

A measuring tool for monitoring critical proximity according to a sixth aspect of the present invention comprises a container made of a hydrogen-containing substance such as polyethylene or acrylic having a thickness equivalent to water of 1 to 5 cm in terms of hydrogen atom density. It is characterized in that neutron activation detection is performed by filling with a Na-containing substance.

According to the present invention, a simple neutron measurement monitoring jig is periodically sampled to obtain a schematic change with time of the neutron flux, so that critical proximity monitoring can be performed in a batch-like manner. In the unlikely event that a neutron emission accident occurs, the subsequent data change can be grasped by continuously sampling.

Since Na is commonly used as salt and the like, and the measurement of radiation is easy, it is possible to measure the amount of neutron irradiation at extremely low cost. By adjusting the thickness of the container containing the hydrogen-containing substance or arranging a thermal neutron absorber on the outer periphery, the energy response characteristics with respect to incident neutrons can also be adjusted.

According to a seventh aspect of the present invention, there is provided a neutron measurement for activation detection for obtaining a general change with time of a neutron flux by combining a plurality of radionuclides having different half-lives generated by irradiation with neutrons. A jig is attached to the object to be monitored, and sampling is performed periodically.

According to the present invention, a simple neutron measurement jig is periodically sampled to obtain a schematic change with time of the neutron flux, so that not only a critical proximity monitoring method can be obtained, but also a method of monitoring critical proximity. Even in the event of a neutron emission accident, it is possible to grasp changes in the data after the fact by sampling.

In the measuring jig for critical proximity monitoring according to claim 8, the plurality of radionuclides having different half-lives generated by irradiating neutrons are In116m, Eu152m, Na24, La140,
It is characterized in that neutron activation detection is performed by including at least two kinds of Au198.

The half-life of these radionuclides is 54.
2 minutes, Eu152m for 9.3 hours, Na24 for 15 hours, La140
For 40.3 hours and Au198 for 64.8 hours (2.7 days). All of them are chemically stable and harmless, and radiation measurement is easy. Considering the case where neutrons are irradiated for a long time, those with a short half-life mainly have the information immediately before the end of irradiation, and those with a long half-life have the entire information. It is possible to estimate the neutron irradiation dose for each section by assuming a plurality of time sections.

[0025]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS First, a description will be given of a method for monitoring how a system including a fuel region is approaching criticality, which is a premise of each embodiment according to the present invention.

The neutron flux φ for a subcritical fuel system including a neutron source can be explained by a single point reactor model. Assuming that the effective neutron multiplication factor is k and the proportional constant is b, “φ = b /
It is widely known that it can be expressed in the form of (1-k).

However, the neutron flux φ when the neutron source is arranged inside or on the side of the fuel region is generally expressed as “φ = b
/ (1-k) ", and the present inventors have found that a very simple empirical formula" φ = a + b / (1-k) "is very useful. Numerical Technology vol.97, p.131
(1992) "Active Neutron Multiplication Method for
Fuel Lattices in Water ". This empirical formula is a very simple approximation simply by adding the constant a, but the harmonic component of the large neutron flux excited by the neutron source exists. It has proven to be an excellent empirical formula that can be applied regardless.

In the present invention, this empirical formula is used as a basic formula.

A neutron flux φ based on a spontaneous neutron source distributed inside the fuel region or an artificially placed external neutron source
And a a _0, b _0, a, and b constants, composition of the fuel, the concentration and shape constant one fuel region of thinking scheme effective multiplication factor is different, with respect to systems large effective multiplication factor k ₀ Where φ ₀ = a ₀ + b ₀ / (1−k ₀ ), φ = a + b / (1−k) for a system with a small effective multiplication factor k, and R = k / k ₀ , ρ ₀ = (1−k ₀ ) / k ₀ .

It has been confirmed by numerical calculation that the value of R is constant without depending on k ₀ even in a system relatively close to criticality. Therefore, the value is obtained by calculation or the like, and φ ₀ −a ₀ = φ ₀ (1−a ₀ / φ ₀ ) = b ₀ / (1−k ₀ ), φ−a = φ (1−a / φ) = B / (1−Rk ₀ ) = {b / (1−k ₀ )} / [1+ (1−R) / ρ ₀ ] to obtain the following equation (φ ₀ / φ) = (b ₀ / b) obtaining a · {(1-a / φ ) / (1-a 0 / φ 0)} × [1+ (1-R) / ρ 0]. (A / φ) and (a)
₀ / φ ₀ ) is sufficiently smaller than 1, and the neutron flux ratio (φ ₀ / φ) has a substantially linear relationship with (1 / ρ ₀ ).

Here, the relationship between the neutron flux ratio (φ ₀ / φ) and (1 / ρ ₀ ) is determined in advance, and this characteristic is used. That is, φ ₀ and φ are obtained by measurement, and the above property is used to calculate (1 / ρ) from the neutron flux ratio (φ ₀ / φ) value.
₀ ) to determine k ₀ .

In the fuel region arranged in the neutron non-moderator space, a system having a large effective gain k ₀ is a system in which a neutron reflector is mounted on the side of the fuel region, and a system having a small effective gain k is This system does not have a neutron reflector on the side of the fuel area.

The system having a large effective multiplication factor k ₀ in the fuel region arranged in the neutron moderator space is a system in which the neutron absorber is not mounted on the side of the fuel region, and the system having a small effective multiplication factor k is a neutron at the side of the fuel region. It is a system equipped with an absorber.

Next, the parameter ratios (a / b) and (a
_A method of specifically searching for a place where the absolute value of ( ₀ / b ₀ ) becomes small will be described.

The effective multiplication factor k can be obtained by using the eigenvalue calculation method for one type of fuel concentration system. The neutron flux φ is obtained by calculating the multiplied neutron flux (absolute value) accompanying the neutron source arrangement.

Here, the basic expression is "φ = a + b / (1-
k) ", the unknowns are a and b. Therefore, the above-mentioned k and φ can be obtained for two types of fuel systems having different fuel concentrations, simultaneous equations can be made, and the unknowns a and b can be determined.

Looking at these characteristics from the calculation results of a and b, a becomes very large in the vicinity of the neutron source as compared with b, so that it is not suitable for accurately obtaining the influence of multiplication. a
Can be negative.

By calculating for three or more concentrations,
The applicability of the basic formula to the target condition can be examined.

When the size of the fuel region is substantially the same as the size of the light water reactor fuel assembly, (a / b) and (a ₀ / b)
_The location where ( ₀ ) is sufficiently smaller than 1 is on the opposite side of or near the neutron source with respect to the fuel region.

In order to further limit, a location is surveyed in a direction parallel to the axis of the fuel region, centering on a position where the axis of the fuel region is perpendicular to the straight line connecting the neutron source and the center of the neutron detector. This survey is performed in three dimensions, and a place where (a / b) and (a ₀ / b ₀ ) are sufficiently smaller than 1 is determined.

Numerical calculations were performed to quantitatively demonstrate the above theory. Next, embodiments of the present invention will be described with reference to the drawings.

(First Embodiment) FIG. 1 is a calculation model of a simulated air system. In the fuel region 2 arranged in the neutron non-moderator space 1 (space simulating in air),
Water 3 as a neutron moderator is arranged only in the fuel region 2, and dilute water having a density of 1/100 is arranged on the outer periphery of the fuel region 2 so that neutron diffusion calculation can be performed while simulating air approximately. This is a simulated air system, and shows a case where water of the neutron reflector 4 is arranged so as to sandwich the fuel region 2 and a case where it is not arranged.

The external neutron source 5 and the neutron detector 6 are both surrounded by water, which reduces the influence of the attachment / detachment of the reflector and prevents the occurrence of a calculation error in the diffusion calculation. The thickness of the surrounding water or hydrogen-containing substance is preferably at least 3 cm in terms of water, and preferably the thickness up to the vicinity of the detachable reflector 4 if possible.

The fuel region 2 has a cross section of 21 × 21 pitch in which fuel rods 7 having a diameter of 1.2 cm and filled with UO ₂ pellets having a diameter of 1 cm are arranged in a square at a lattice pitch of 1.52 cm. In order not to complicate the calculation, an XY two-dimensional system (that is, infinite length in the axial direction) was used.

The neutron effective multiplication factor of the system was changed by changing the U235 enrichment of the UO ₂ pellet from 1% to 5%.

The calculation here is carried out by calculating an effective multiplication factor by eigenvalue calculation without disposing a neutron source, an artificial external neutral source 5
Neutron flux calculation (fixed neutron source calculation based on an external neutron source, hereinafter referred to as “external neutron source multiplication calculation”) in which the external neutron source 5 is not disposed and the fuel region 2
A neutron flux calculation (a fixed neutron source calculation based on a spontaneous neutron source, hereinafter referred to as a "multiplication calculation of a spontaneous neutron source") was performed when a uniform spontaneous neutron source was distributed throughout.

FIG. 2 shows underwater calculations when a Cd plate 9 as a neutron absorber is not arranged on the side surface of a fuel region 2 arranged in a neutron moderator space 8 using water 3 as a neutron moderator. It is a system.

The difference between the fuel region 2 and the case of FIG. 1 is that the underwater system has a large effective multiplication factor.
It is slightly smaller so as to have a cross section of 19 × 17 pitch instead of 1 × 21. The other points are the same as those in FIG.

FIG. 3 shows the calculated values for the system of FIG.
FIG. 4 shows the calculated values for the system of FIG.

The horizontal axis shows the reciprocal of the reactivity (ρ ₀ = (1−k ₀ ) / k ₀ ) focused on in the present invention on the lower side, the effective multiplication factor k ₀ on the upper side, and the vertical axis shows the present invention. Indicates the neutron flux ratio of interest.

More specifically, in FIG. 3, the horizontal axis represents the effective multiplication factor kR when the reflector 4 is mounted, and the vertical axis represents the reflector for the neutron flux φN when the reflector 4 is not mounted. 4 and the ratio of neutron flux φR (φR / φ
N). In FIG. 4, the horizontal axis represents the effective multiplication factor kN when the neutron absorber (Cd) 9 is not mounted, and the vertical axis represents the neutron absorber (Cd) for the neutron flux φC when the neutron absorber (Cd) 9 is disposed. Cd) The ratio (φN / φC) of the neutron flux φN when 9 is not arranged.

In each of the figures, two straight lines are shown, the solid line is a case where the external neutron source 5 is mounted (multiplication of the external neutron source), and the broken line is a case where the external neutron source 5 is removed and one line is added to the fuel region. Is used as a neutron source (spontaneous neutron source multiplication).

When a ₀ , b ₀ , a, b are obtained from the respective straight lines, the value of a ₀ is positive in the example of FIG. 3, and the ratio of the constants (a ₀ / b ₀ ) Is a value whose absolute value is much smaller than 1. In the case of FIG. 4, a ₀ is negative, and (a ₀ / b ₀ ) is a relatively large value whose absolute value cannot be ignored compared to 1.

When the fuel assembly is arranged in water, a ₀ is positive for a fuel assembly of a boiling water reactor (about 14 cm × 14 cm in cross section), and is a positive _value for a fuel assembly of a pressurized water reactor (about 21 cm in section).
(cm × 21 cm), it is slightly negative, and it is known from experiments and numerical calculations that the negative absolute value tends to increase as the fuel assembly increases. This allows
It is clear that a0 (cross section about 52 cm × 46 cm) in FIG. 4 is negative.

Under the above conditions, the linearity is good in both cases of FIGS. 3 and 4, but there is a difference between the external neutron source multiplication and the spontaneous neutron source multiplication.

Further, looking at the figure in detail, in the case of FIG. 3, the effective gain is an excellent straight line in a wide range from below 0.8 to 0.97, but in the case of FIG.
When it is smaller than around 0.90, it can be seen that it is slightly deviated from the straight line.

Such a characteristic is represented by a constant ratio (a ₀ / b 0)
Is determined by whether or not the absolute value of the ratio is smaller than 1. Therefore, it is preferable to reduce the absolute value of this ratio as much as possible.

By examining the spatial distribution of the ratio of the neutron flux in the system with the reflector 4 to the system without the reflector 4 in the system of FIG.
In the example of the system of FIG. 2, it is constant. This is because in the system shown in FIG. 1, although the water moderator exists around the neutron detector 6, the water moderator is not entirely extended to the reflector to be attached / detached. That is, there is a space in which water is cut off between the water around the neutron detector 6 and the detachable water reflector 4. Therefore, it is desirable to arrange a hydrogen-containing substance also in this space.

Therefore, a neutron detector moderator having a thickness of at least 3 cm in water equivalent is disposed on the neutron reflector 4 attaching / detaching side surrounding the neutron detector 6 or on the neutron absorbing body 9 attaching / detaching side, and the neutron reflector attaching / detaching or neutron absorbing It is preferable to adopt a configuration in which the influence of a change in the neutron field due to body attachment / detachment is suppressed, and the thickness is set so as to approach the neutron reflector 4 or the neutron absorber 9.

The neutron detector 6 and the neutron source 5, at least the water 3 surrounding the neutron detector 6 or the thickness of the hydrogen-containing substance is 3 cm or more in terms of water.
Alternatively, if the thickness is set so as to be close to the neutron absorber 9, the effect of attachment / detachment of the reflector 6 or the absorber 9 is reduced by the neutron deceleration and absorption effect of hydrogen, and the constant (a ₀ , a) appearing in the basic formula , And (b ₀ , b) can be reduced. Therefore, the calculation errors of these constants obtained by the calculation are also reduced, and the influence thereof can be reduced.

(Second Embodiment) Next, in the event that a criticality accident occurs and neutron emission continues, the neutron irradiation dose (time integral of neutron flux) is measured at a simple, easy and extremely low cost. The possible measuring method and measuring jig will be described.

The first measuring jig is characterized in that a container of a hydrogen-containing substance such as polyethylene or acrylic having a thickness of 1 to 5 cm in water equivalent is filled with a sodium-Na-containing substance such as salt and soda glass. This is a neutron measurement jig for detecting neutron activation. Since Na is common and easy to measure as salt or the like, it is possible to measure the amount of neutron irradiation at extremely low cost.

The second measuring jig is characterized in that a plurality of radionuclides having different half-lives generated by irradiating neutrons are combined to obtain a schematic temporal change of a neutron flux. It is a neutron measurement jig.

More specifically, as a plurality of radionuclides having different half-lives generated by irradiation with neutrons, In116m, E
u152m, Na24, La140, Au198 at least 2
It is characterized by selecting species.

The half-life of these radionuclides is In 116m
54.2 minutes, Eu152m for 9.3 hours, Na24 for 15 hours, L
a140 is 40.3 hours and Au198 is 64.8 hours (2.7 days). All are chemically stable and harmless, and radiation measurement is easy. Considering the case where neutrons are irradiated for a long time, those with a short half-life mainly have the information immediately before the end of irradiation, and those with a long half-life have the entire information. It is possible to estimate the neutron irradiation dose for each section by assuming a plurality of time sections.

That is, the irradiation time is divided into a plurality of sections within 5 sections in accordance with the number of nuclides to be used, and the expression of the sum of the amounts of radioactivity taking into account the activation rate and the decay rate at the sectioned time is given by Create for nuclides, set simultaneous equations,
If the neutron flux for each time section is solved as an unknown value, the value of the neutron flux for each time section (that is, the time-dependent change of the neutron flux)
Is obtained. Although the contents of the simultaneous equations are simple but complicated, they are not described, but the expert can easily create the simultaneous equations by the above description.

Although five typical types have been described above, the present invention is not limited to these. While many can be selected, for example, MnMn55, when activated, produces Mn56 with a half-life of 2.56 hours and emits gamma rays such as 0.845 MeV, which are easily measured.

In a criticality accident, it is considered that the neutron counting rate (burst) temporarily becomes high in a short time in the initial stage, and thereafter changes to a low neutron counting rate. In such a case, to roughly track changes in the neutron flux over time,
For example, a combination of In116m, Na24, and Au198 is convenient.

FIGS. 5 (tower type) and FIG. 6 (side-by-side type) show examples of measuring jigs for detecting neutrons by the activation method using the above five kinds of nuclides. Each of them has basically a plug type module configuration, and uses polyethylene (PE) 11 as a neutron moderator. Acrylic is perfectly acceptable.

In FIG. 5A, salt (NaCl) 1
2 is filled, a gold foil 13 is stored underneath, and NaCl is filled inside a PE wrapped with cadmium (Cd) 14 further below. (Cd) 14 absorbs the incident thermal neutrons, and the transmitted fast and epithermal neutrons are decelerated by the internal PE to activate Na in NaCl.

Therefore, in the configuration of FIG. 5A, neutron spectrum information can be obtained at low cost. The configuration most simplified in FIG. 5A is to attach the uppermost portion to the base 15.

FIG. 5B shows a module on which five types of substances to be activated are mounted. Unlike FIG. 5 (A), the purpose is to evaluate the temporal change of the neutron emission rate. (A) and (B) may be combined.

FIG. 6 shows a configuration in which FIG. 5B is arranged side by side. Four types of PE blocks 16 are attached to the base 15. It is substantially the same as FIG.
The PE block 16 at the lower right in FIG. 6A has a configuration in which a gold pillar 13 and an In foil 18 are wound around a PE column 17.

FIGS. 5 and 6 show two types of configurations, but various configurations are conceivable. For example, a skewer configuration may be used. Multiple samples may be stored in a capsule, placed in a pipe with a spacer of about 2 to 5 cm in length, and inserted into a polyethylene block.

At the accident site, it is generally difficult to recover the activated sample immediately after the accident is completed, so it is difficult to apply this method. For the measurement of the integral amount, it is most preferable to use a stainless steel (SUS) or a SUS irradiation amount monitoring test piece which is frequently used for a structural material or the like. SUS
Contains Ni58 and Cr50. Ni can measure fast neutrons, and Cr can measure thermal neutrons.

In the outdoors, it is sufficiently possible to recover the activated sample immediately after the accident.

Therefore, the above-described method and measuring jig described in detail are extremely effective. If the gamma ray level is still high immediately after the accident, a recovery method such as an airship method can be selected according to the location.

[0078]

According to the present invention described above, the criticality monitoring method is carried out to automatically determine whether the system containing nuclear fuel is approaching the criticality and to prevent a criticality accident from occurring. be able to. In addition, it is possible to foresee a criticality accident in advance by grasping changes in the neutron multiplication system by sampling the activation measurement jig periodically, and even if a criticality accident occurs, From the outside of the building, it is possible to estimate the neutron generation rate and its chronological change. Therefore, the present invention relates to (a)
The measuring principle and equipment are simple, (b) the configuration of the measuring instrument and its calibration work are simple, (c) maintenance and management are easy, and (d) it can be provided at extremely low cost.

[Brief description of the drawings]

Fig. 1 Calculation model of simulated airborne system

Fig. 2 Calculation model of underwater system

Fig. 3 Calculation result for simulated air system

Fig. 4 Calculation results for underwater systems

Fig. 5 Tower type neutron measurement jig

FIG. 6 is a side-by-side neutron measurement jig

[Explanation of symbols]

1 ... neutron non-moderator space 2 ... fuel area 3 ... water 4 ...
Neutron reflector, 5 ... external neutron source, 6 ... neutron detector,
7: fuel rod, 8: neutron moderator space, 9: Cd plate, 10
... Plug type module, 11 ... PE, 12 ... Na
Cl, 13: Gold leaf, 14: Cd, 15: Base, 16 ...
PE block, 17 ... PE pillar, 18 ... In foil

 ──────────────────────────────────────────────────続 き Continuing from the front page (72) Inventor Tsukasa Kikuchi 2-1 Ukishima-cho, Kawasaki-ku, Kawasaki-shi, Kanagawa Prefecture F-term in the Toshiba Hamakawasaki Plant (reference) 2G075 AA17 BA03 BA16 CA12 CA38 DA02 DA09 EA01 EA02 EA03 EA04 FA18 FA19 FB10 FB15 FC12 FC13 FC15 FC16 FD06 FD07 GA02 GA09 GA24 GA28 GA34

Claims (8)

[Claims] 1. A neutron detector is disposed on the side or inside of a fuel container, and a neutron flux (φ ₀ ) when the effective multiplication factor is large.
And _{φ 0 = a 0 + b 0} / (1-k 0) the relation between the effective multiplication factor (k ₀₎ the relation between the neutron flux when the effective multiplication factor is small (phi) and the effective multiplication factor (k) phi = A + b / (1−k) (where a ₀ , b ₀ , a, and b are constants) as (φ ₀ /
From the relationship between φ) and φ ₀ , φ, k ₀ , k, (φ ₀ / φ) and (1 / ρ ₀ ) (where ρ ₀ = (1−k ₀ ) / k ₀ ) are almost linear. become seeking region to monitor the change in k ₀ by measuring the phi ₀ and phi, the critical proximity monitoring method k ₀ is equal to or emit an alarm when it exceeds a certain value. 2. In the fuel region arranged in the neutron non-moderator space, the fuel region in the case where the effective multiplication factor is k ₀ is a system in which a neutron reflector is attached to the side of the fuel region, and the effective multiplication factor is 2. The critical proximity monitoring method according to claim 1, wherein the fuel region in the case of k is a system in which a neutron reflector is not mounted on a side surface of the fuel region. 3. In the fuel region arranged in the neutron moderator space, the fuel region in the case where the effective multiplication factor is k ₀ is a system in which a neutron absorber is not attached to the side of the fuel region, and the effective multiplication factor is k. 2. The critical proximity monitoring method according to claim 1, wherein the fuel region in the case of (1) is a system in which a neutron absorber is mounted on a side surface of the fuel region. 4. At least 3 cm in thickness surrounding a neutron detector and on a neutron reflector attaching / detaching side or a neutron absorber attaching / detaching side.
4. The critical proximity monitoring method according to claim 1, wherein a moderator for a neutron detector is arranged. 5. A neutron transport / diffusion calculation for an effective multiplication factor k and a neutron flux φ for at least two types of fuel concentrations is performed on a measured system in which a neutron source is arranged, and a certain value or more except for the vicinity of the neutron source At a distance of
4. A critical proximity device according to claim 1, wherein a location where the absolute values of the ratios (a / b) and (a ₀ / b ₀ ) of the parameters are reduced is set, and a neutron detector is set at the location. Monitoring method. 6. A method for detecting neutron activation by filling a container containing a hydrogen-containing material such as polyethylene or acrylic having a thickness equivalent to water of 1 to 5 cm in water equivalent in terms of hydrogen atom density with a Na-containing material. Measuring jig for critical proximity monitoring. 7. A neutron measurement jig for activation detection for obtaining a schematic temporal change of a neutron flux is attached to a monitoring object by combining a plurality of radionuclides having different half-lives generated by irradiation with neutrons, A critical proximity monitoring method characterized by periodically sampling. 8. The plurality of radionuclides having different half-lives generated by irradiation with neutrons are In116m, Eu152m, N
1. A critical proximity monitoring measuring jig comprising at least two of a24, La140, and Au198, and performing neutron activation detection.