JPH1138147A - Device for detecting neutron flux and temperature in reactor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To easily measure the progresses with time of a neutron flux and irradiation temperature off-line, by providing a detecting mechanism, which is composed of a wire paying-off machine, etc., and in which a wire detecting medium passes through the wire neutron irradiating section between a paying-out container and a housing container and a driving mechanism. SOLUTION: The device for detecting neutron flux and temperature in reactor is constituted of a detecting block 21, which is inserted into a reactor together with a sample to be irradiated, a detecting block maintenance device 22, which performs the maintenance of the block 21 outside the reactor, and an automatic activated sample measuring instrument 23, which measures the radioactivity of a detecting medium. The detecting block 21 is composed of a detecting mechanism 26 and a driving mechanism 27. After irradiation, the detecting mechanism 26 is taken out from the detecting block 21 by using the maintenance device 22. A wire in the detecting mechanism 26 is measured by means of a gamma-ray spectrometer 29 with collimator while the wire is fed at a fixed speed by means of an automatic wire feeding device 28 and neutrons are found. The irradiation temperature is found based on the radio of the activated radioactivity of a wire detecting medium for measuring neutron to that of a wire detecting medium for measuring temperature.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an atom for capsule instrumentation for measuring a time distribution of a thermal neutron flux, a fast neutron flux and a temperature during irradiation in a test or the like in which a material is irradiated with neutrons in a nuclear reactor. The present invention relates to a neutron flux / temperature detector in a reactor.

[0002]

2. Description of the Related Art Conventionally, neutron detectors used for capsule instrumentation when performing a test of irradiating a material with neutrons in a nuclear reactor include a fission counter, a ¹⁰ B counter, and a BF ₃ counter.
Neutron detectors such as counter tubes have been used. Thermocouples and resistance thermometers have been used as irradiation temperature detectors. However, when using a neutron counter or temperature sensor,
It is necessary to use a signal cable or a weak signal cable to which a high voltage is applied, which makes it difficult to handle the irradiation capsule in the reactor and limits the insertion conditions. Further, when the temperature inside the reactor becomes high, there is a problem of heat resistance, and thus usable detectors are limited.

For this reason, activation analysis has been used as a method for easily obtaining a neutron flux at an irradiation location inside a nuclear reactor even at high temperatures. Metal foils, metal wires, metal balls, etc. with a relatively large neutron capture cross section or nuclear reaction cross section for thermal neutrons or fast neutrons are inserted together with the irradiated object, and after irradiation, the total The number of neutrons can be obtained. However, it was impossible to measure the time change of the neutron flux with this method.

In order to solve this drawback, as shown in FIG. 1, the above-mentioned metal wire, metal ball, etc. are sent to an irradiation place at regular intervals using an introduced pipe, and neutrons are analyzed by activation analysis. An automatic activation measurement device for measuring the time distribution of the bundle has been devised.

[0005] However, it is necessary to install a ball introduction tube, a delivery device, and the like, and the device is large-scale.

[0006]

When the above-mentioned counter tube is used, a through hole for a signal cable or the like is required, and a troublesome operation for managing the reactor for handling the cable or the like is required. Needed. In addition, when the reactor is a high-temperature gas reactor, a fast reactor, or the like, it has been very difficult to manufacture a high-temperature counter tube in a high-temperature primary cooling environment.

In general, when the activation analysis used in the irradiation test is used, it is not possible to determine the passage of time. In addition, the above-mentioned automatic activation measurement apparatus has a drawback that the introduction pipe or the like becomes an obstacle similarly to the cable or the like, and the introduction pipe or the delivery apparatus and the like need to be installed, so that the apparatus becomes large-scale.

[0008] Therefore, an object of the present invention is to dispose a small detection block near an irradiation sample without using a signal cable, a power supply cable, an introduction pipe, and the like.
An object of the present invention is to provide a neutron flux / temperature detecting device in a nuclear reactor that easily and offline measures the time lapse of thermal neutron flux, fast neutron flux and irradiation temperature.

[0009]

The present invention provides a thermal neutron flux,
The present invention relates to a neutron flux / temperature detector in a reactor for measuring a time distribution of a fast neutron flux and a temperature, wherein the detector comprises: (a) a wire feeding container surrounded by a neutron absorber or a neutron moderator; , A wire detection medium containing an activated material having a thermal neutron capture cross section stored therein, a mechanism for feeding out the detection medium, and a thermal neutron absorbing material for containing the fed wire detection medium. A detection mechanism having a structure in which a wire detection medium is passed through a wire neutron irradiation section between the feeding container and the storage container, and (b) a mechanical source such as a spring or a spring is used as a power source. And a drive mechanism for feeding and storing the wire detection medium in the wire feeding container at a constant speed.

[0010] In particular, when the time distribution of the irradiation temperature is obtained, the temperature of the same or different content of the same activated material is determined together with the wire detection medium containing the activated material having a thermal neutron capture cross section. A structure in which the measurement wire detection medium is fed out at the same speed, a thermal neutron collimator made of a material with a large thermal neutron capture cross section is installed in front of the temperature measurement wire detection medium, and the degree of opening of the collimator window Can be performed by adding a mechanism that changes according to the temperature of the bimetal or the like.

[0011]

DESCRIPTION OF THE PREFERRED EMBODIMENTS As shown in FIG. 2, the apparatus of the present invention comprises a detection block 21 inserted into a nuclear reactor together with an irradiation sample, a detection block maintenance device 22 for performing maintenance of the detection block outside the reactor, and a detection medium. And an automatic activated sample measuring device 23 for measuring the radioactivity of the sample.

As shown in FIG. 3 or FIG. 5, the detection block includes a neutron absorbing material 31, a resonance neutron absorbing material 32, or a wire feeding container 33 surrounded by a neutron moderator, and a neutron capture device contained therein. A wire detection medium 34 containing an activated material having a cross-sectional area, a mechanism 35 for feeding the detection medium to the outside, a wire storage container 36 for storing the fed wire detection medium, and a wire winding for storing the detection medium inside the container It comprises a mechanism 37 and has a structure in which a wire detection medium is activated in a wire neutron irradiation section 38 between the feeding container and the storage container. In addition, the influence of the background due to neutrons irradiated when the wire detection medium is present in the feeding container and the storage container is reduced by surrounding the periphery with a neutron absorbing material or a neutron moderator.

The effect of the background is that the wire detection medium wound up in the storage container for the detection medium cannot be completely shielded and is activated.
Therefore, the total amount of the activated wire detection medium is the amount activated in the feeding mechanism and the storage container and the amount activated in the neutron irradiation section provided between the feeding container and the storage container. It is sum. For this reason, the value obtained by subtracting the activated amount of the wire detection medium left in the feeding container to the end as the background amount is the pure activation amount of the wire detection medium, and the thermal neutron flux is corrected. . As a result, since the position of the wound wire corresponds to the time of activation in the neutron irradiation section, the time distribution of thermal neutrons (thermal neutron amount corresponding to the measurement time) can be measured.

When the wire detection medium has a large resonance absorption region for neutrons, if the medium is wound up in a feeding container and a storage container, when the neutrons are irradiated with the large resonance absorption region, the radiation is radiated. Into a background. In order to prevent this, if the same material as the wire detection medium is fed out and used as a shielding material for the storage container, activation of the resonance absorption region can be prevented most efficiently.

After the irradiation, as shown in FIG. 2, the detection mechanism 26 is taken out of the detection block 21 by using a detection block maintenance device 22 including a detection medium removing / mounting mechanism 24 and a mainspring set mechanism 25 outside the reactor. Alternatively, the wire detection medium may be used as it is as an activation sample, or the wire detection medium may be taken out of the detection mechanism once, and set again in a newly-produced measurement feeding / storing mechanism. When the drive mechanism 27 can be reused, the drive mechanism 27 sets a mainspring or the like to a specified amount.

The wire in the detecting mechanism 26 or the wire set in the measuring feeding / storing mechanism is measured at a constant speed by an automatic wire feeder 28 or at regular intervals by a gamma ray spectrometer device 29 with a collimator. Determine the amount of radioactivity. The neutron flux is calculated based on this radioactivity. At this time, the background is subtracted based on the activation amount of the wire left in the feeding container. With the above operation, the time distribution of the thermal neutron flux or the fast neutron flux is measured with high accuracy.

When measuring the high-speed neutron flux measurement, as shown in FIG. 5, a wire feeding container and a wire storage container which are surrounded by a thermal neutron absorbing material and a high-speed neutron moderator are used.

In order to obtain the irradiation temperature, as shown in FIG. 6, together with a wire detection medium for neutron flux measurement, a wire detection medium 61 for temperature measurement having the same content or different content of the same activated material is used. A thermal neutron collimator 6 made of a material having a large thermal neutron capture cross-section is provided in front of the temperature detecting wire detection medium.
2, a time distribution of the irradiation temperature is obtained by adding a mechanism for changing the degree of opening of the window of the collimator depending on the temperature of the bimetal 63 or the like.
At this time, the irradiation temperature can be determined based on the ratio between the neutron measurement wire detection medium and the activation radioactivity of the temperature measurement wire detection medium.

That is, when the window of the collimator is opened, the amount of neutron flux irradiated on the wire detection medium through the collimator increases, and when the window is closed, the amount of neutron flux irradiated decreases. Therefore, if the reference thermal neutron flux is measured with the other wire detection medium and the other thermal neutron flux measured according to the collimator opening is calibrated,
You can see how much the collimator was open. Since the degree of opening of the collimator depends on the temperature due to the bimetal,
The irradiation temperature can be determined. Therefore, the time distribution of the irradiation temperature (the temperature corresponding to the measurement time) is obtained by calculating the activation amount at the same position of the two wire detection media.
Can be measured.

For driving a wire or the like, a heat-resistant and small driving mechanism for feeding and storing the wire or the like at a constant speed using mechanical stress such as a spring or a spring as a driving source is used, and a driving or signal cable or the like is used. It can be omitted.

Further, in the present invention, a mechanical source such as a mainspring or a spring is used as a driving source for feeding and storing the detection medium, and no external power source is required.
It is possible to omit driving or signal cables and the like. By surrounding the detection medium with a large neutron absorption cross-sectional area in the feeding section and the storage section, the amount of activation until irradiation is reduced as much as possible, and background correction is performed using the remaining detection medium without feeding, and thermal It is characterized in that the amount of elapsed time of neutrons and fast neutrons is accurately obtained. The irradiation temperature is determined by reheating the irradiation temperature to a change in neutron flux based on the principle of a bimetal thermometer.

Further, an embodiment of a driving mechanism for driving the above detection mechanism will be described with reference to FIG. As a material of the drive mechanism, a high-temperature stainless steel material or a high-temperature ceramic material is used for the gears and the like and their support blocks in consideration of activation of metal. Further, as a method of performing constant wire driving, a mechanical mainspring clock-type constant speed maintaining mechanism having a structure similar to a mechanical timepiece conventionally used is used. By this maintenance mechanism, the rotating body for feeding the wire and the rotating body for winding the wire are driven to keep the speed for feeding and storing the wire constant. As a driving source, a mainspring made of a heat-resistant steel A-286 material (heat-resistant temperature: 650 ° C. or more) that can be used even at a high temperature is used. With such a configuration, driving at 650 ° C. becomes possible.

Further, in the present driving mechanism, mechanical stress is used as a driving source. However, when a power cable or the like can be easily laid in a nuclear reactor, driving by a power supply is also possible.

[0024]

[Example 1] [Measurement of thermal neutron flux] (Measurement in the case where the thermal neutron flux is large:
Neutrons having energy not less than thermal neutrons are present in the reactor, but this is applicable when there are few neutrons.) The thermal neutron flux is measured in the reactor neutron flux / temperature detector according to the embodiment of the present invention. The case will be described with reference to FIG.

As the thermal neutron activation material used for obtaining the thermal neutron flux, metal Co was used, and ⁵⁹ Co (n,
gamma) using ⁶⁰ Co reaction. ⁵⁹ Co is the presence of 100%, the neutron capture cross section is 37.2Barn.

Stainless steel is used as a base material of the wire as a detection medium, and 1% of Co is contained therein. The thickness of the wire is 0.5 mm in diameter. The wire is unreeled every lcm every 10 hours to measure the thermal neutron flux, and the thermal neutron flux is measured for 10 days. The wire neutron irradiation section 38 between the feeding container 33 and the storage container 36 is 1 cm. The required wires are 24
24 cm as a wire for 0 hour, 1 cm as a wire for background evaluation and 4 as a wire for winding
cm and the total amount is 30 cm. ⁵⁹ Co atomic number N in lcm detection medium at this time (0) N (0) = 6.022 × l0 23 × (3.14 × 0.05
× 0.05 × 1 × 7.9I × 0.01) /55.8=
A 6.7 × l0 ^18.

On the other hand, boron nitride (BN) is used as a thermal neutron absorber for the payout portion and the storage portion where the wires are stored. The thermal neutron absorption coefficient by the thermal neutron absorber can be calculated by the following equation. _{D = exp (-Σ m · t} a) where sigma _m is macro capture cross section thermal neutron (0.0253eV). In the case of BN (boron nitride), this Σ _m
Is 16.15 cm ^-1 . FIG. 4 shows the thermal neutron decay rate when the thickness of BN is changed from 0 cm to 0.3 cm based on this value. From this result, if there is a BN thickness of 0.3 cm, the thermal neutrons inside the feeding part container are 1000
It will be a fraction.

On the other hand, a general formula relating to the amount D of radioactivity activated by thermal neutrons is expressed as follows. D (t ₀ , t _l , t ₂ ) = N (0) (σφ / λ) (1-e
xp (−λt ₀ )) (exp (−λt ₁ ) −exp (−λ
(T _l + t ₂₎₎ where, D: number of collapsed was measured from the radiation of radioactivity t _o: time of irradiating the detection medium (sec) t _l: time after the detection medium irradiated until radioactivity (Sec) t ₂ : Measurement time of radioactivity (sec) N (0): Number of atoms in detection medium σ: Neutral capture cross section (cm ² ) λ: Decay constant of activated nuclide in detection medium φ: Neutron flux (1 / cm ² · sec).

The decay number D for the radioactivity of the detection medium 1 cm and the radioactivity of the background correction wire 1 cm is calculated based on this equation. As a calculation condition, the neutron flux φ in the irradiation field in the reactor is averaged 1 × 10 ¹² / cm ² · s
ec. ^{59 37.2ba} the neutron capture cross-sectional area of the Co
rn. In addition, the period from irradiation to measurement of radioactivity is one day, and the measurement time is 1000 seconds.

For the radioactivity of 1 cm of the detection medium, ⁶⁰ C
Considering the long half-life of o, calculate for the first 10 hours of irradiation. The parameters at this time are as follows. t _o : 36000 sec (10 hours) t _l : 864000 sec (10 days) t ₂ : 1000 sec N (0): 6.7 × 10 ¹⁸ pieces σ: 37.2 barn λ: 4.17 × 10 ^-9 φ: 1 × When a parameter of 10 ¹² / cm ² sec or more is input to the equation and calculated, the decay number D is 1
When converted per second, it is 37400 / sec.

On the other hand, the background correction wire lc
The parameters for calculating the decay number D for m radioactivity are as follows: t ₀ : 864000 sec (10 days) t _l : 864000 sec (1 day) t ₂ : 1000 sec N (0): 6.7 × 10 ¹⁸ σ: 37.2 barn λ: 2.20 × 10 ^-8

The neutron flux is φ: l × 10 ⁹ / cm ^{2 in} consideration of the attenuation factor of 0.001 due to boron nitride.
-It becomes sec. When the above parameters are input into the formula and calculated, the number of decay D is 896 / sec when converted per second.
Becomes Since 23/24 of this value is the background, the actual number of decay is 37404- (23
/ 24) × 896 = 36546, which indicates that the background contribution is 2.3%.

Next, it is confirmed that the number of decay is counted without any problem by the apparatus for measuring the radioactivity of the detection medium.
The Ge gamma ray spectrometer shown in FIG. 2 is used as a detector for counting. As the Ge detector, one having a relative detection efficiency of 20% for 1.33 MeV gamma rays of ⁶⁰ Co is used. In this case, about 20% of the gamma rays incident on the detector are gamma ray photo peaks to be counted. A collimator having a diameter of 1 cm is used, and the distance between the wire and the Ge detector is 10 cm. The counting time is 1000 seconds as described above. The count value Na when the detection medium obtained at this time is counted for 1000 seconds.
N _a = 37404 × is (3.14 × 0.5 × 0.5) /
(4 × 3.14 × 10 × 10) × 0.2 × 1000 = 4
678. On the other hand, the count value N _b for the background correction wire is N _b = 869 × (3.14 × 0.5 × 0.5) (4 ×
3.14 × 10 × 10) × 0.2 × 1000 = 112.

These values are sufficient count values for accurately obtaining the thermal neutron flux. Further, as described above, boron nitride (heat-resistant temperature: 900 ° C. or more) as a thermal neutron absorber, and stainless steel (heat-resistant temperature: 100 ° C.) as a detection medium.
(0 ° C. or higher), etc., makes it possible to measure the time distribution of the thermal neutron flux at an irradiation temperature up to 900 ° C.

[0035]

Example 2 (Measurement of thermal neutron flux in a field including a neutron flux in the resonance neutron energy region) Here, an example in which the detection medium has a resonance neutron absorption cross section is shown in FIG. This will be described with reference to FIG. As described above, when metal Co is used as the activation material for thermal neutrons and the ⁵⁹ Co (n, γ) ⁶⁰ Co reaction is used, a large resonant neutron capture cross section at 0.132 keV outside thermal neutrons is obtained. The peak shows 7000 barn and the peak width is about 100 eV. If the neutron flux to be measured is large in this region, metal Co or an alloy containing metal Co is added as a shielding material in addition to the above-described boron nitride as a material surrounding the feeding container and the storage container.

The neutron absorption coefficient of the Co absorber can be calculated by the following equation. _{D = exp (-Σ m · t} a) where sigma _m is macro capture cross section of neutrons (0.132keV). In the case of this Co resonance neutron capture region,
The sigma _m is 597cm ^-1. Based on this value, when the same stainless steel as the activation detection medium containing 1% Co is used and its thickness is set to 1 cm, the neutron decay rate in this region is calculated as follows: D = exp (-597 * 0. 01 * 1) = 0.0025, and the neutron flux in this resonance region can be sufficiently reduced to 0.25%.

In the case of performing temperature measurement, the same material as the activation detection material is added to the collimator material of the thermal neutron collimator made of a material having a large thermal neutron capture cross section on the front surface of the wire detection medium by the same method. In addition, the neutron flux in the resonance region serving as the background can be sufficiently reduced.

[0038]

Third Embodiment [Measurement of Fast Neutron Flux] A case of measuring a fast neutron flux in the reactor neutron flux / temperature detector according to the embodiment of the present invention will be described with reference to FIG.

As a fast neutron activating material used for obtaining a fast neutron flux, metallic Ni is used and ⁵⁸ Ni (n,
p) ⁵⁸ using a Co reaction. ⁵⁸ Co has an abundance of 67.8%, a threshold energy of 2.9 MeV, and 5 Me
The nuclear reaction cross section for V neutrons is 0.51 barn
It is.

Assuming that stainless steel is used as a base material of the wire serving as a detection medium and that 15% is included as natural nickel Ni, the content of ⁵⁸ Ni (67% of natural Ni) is 10%. %. The thickness of this wire is 0.5 mm in diameter. Put this wire in 10 hours
cm and measure the fast neutron flux.
Measure the fast neutron flux for 10 days. The wire neutron irradiation section between the feeding container and the storage container is 1 cm. The required wire is 24c as a 240 hour wire
m, lcm as a wire for background evaluation, and 4 cm as a wire for winding, the total amount is 30 cm
And The number of atoms N of ⁵⁸ Ni in lcm detection medium at this time
(0) N (0) = 6.022 × l0 23 × (3.14 × 0.05
X 0.05 x I x 7.9 l x 0.1) 58.7 = 6.3.
7 × 10 ¹⁹ is obtained.

On the other hand, the general formula for the amount D of radioactivity activated by fast neutrons is expressed as follows. D (t ₀ , t _l , t ₂ ) = N (0) (σφ / λ) (1-e
xp (−λt ₀ )) (exp (−λt ₁ −exp (−λ
(T _l + t ₂₎₎ where, D: number of collapsed was measured from the radiation of radioactivity t _o: time of irradiating the detection medium (sec) t _l: time after the detection medium irradiated until radioactivity (Sec) t ₂ : measurement time of radioactivity (sec) N (0): number of atoms in the detection medium σ: neutron capture cross section (cm ² ) λ: decay constant of activated nuclide in the detection medium φ: neutron flux ( 1 / cm ² · sec).

Based on this equation, the decay number D for the radioactivity of the detection medium 1 cm is calculated. As calculation conditions, the neutron flux φ in the irradiation field in the nuclear reactor was averaged 1 × 10 ¹² / cm ²
・ Set to sec. The neutron nuclear reaction cross section of ⁵⁸ Ni is 0.5
1 barn. In addition, the period from irradiation to measurement of radioactivity is one day, and the measurement time is 1000 seconds.

For the radioactivity of 1 cm of the detection medium, ⁵⁸ C
Considering the relatively long half-life of o, calculate for the first 10 hours of irradiation. The parameters at this time are as follows. t _o : 36000 sec (10 hours) t _l : 864000 sec (10 days) t ₂ : 1000 sec N (0): 6.37 × 10 ¹⁹ pieces σ: 0.51 barn λ: 1.13 × 10 ^-7 or more parameters When input into the equation and calculated, the number of collapses D is 120360 / sec when converted per second.

If carbon (C) having a thickness of 5 cm or more is used as the neutron moderator and the structure surrounding the wire feed-out part and the storage part is used, the fast neutrons have a threshold energy of 2.9 MeV or less. Can reduce background effects. In addition, the same boron nitride (BN) as described in the thermal neutron measurement method above was used.
By surrounding with 3 cm, the influence of background nuclides due to thermal neutron capture, which is the background of the wire, can be reduced.

Next, it is confirmed that the number of decay can be counted without any problem by the apparatus for measuring the radioactivity of the detection medium.
The Ge gamma ray spectrometer shown in FIG. 2 is used as a detector for counting. As the Ge detector, one having a relative detection efficiency of 20% for 1.33 MeV gamma rays of ⁶⁰ Co is used. In this case, it was incident on the detector
About 30% of the 0.81 MeV gamma ray of ⁵⁸ Co is a gamma ray photo peak to be counted. Lcm in diameter
And the distance between the wire and the Ge detector is 10 cm. The counting time is 1000 seconds as described above. 1000 of the detection medium obtained at this time
The count value Na when counting seconds is Na = 120360 × (3.14 × 0.5 × 0.5)
(4 × 3.14 × 10 × 10) × 0.3 × 1000 = 2
2579. This value is a sufficient count value for accurately obtaining a fast neutron flux.

As described above, carbon (heat-resistant temperature: 1000 ° C. or more) as a neutron moderator, boron nitride (heat-resistant temperature: 900 ° C. or more) as a thermal neutron absorber, and stainless steel (heat-resistant temperature: 1000 ° C.) as a detection medium ℃ or more) makes it possible to measure the time distribution of the fast neutron flux at an irradiation temperature up to 900 ° C.

[0047]

Fourth Embodiment [Case of Measuring Temperature] A case of measuring an irradiation temperature in a neutron flux / temperature detector in a nuclear reactor according to an embodiment of the present invention will be described with reference to FIG.

Along with the wire detection medium containing the neutron capture cross-section activated material used in the thermal neutron detection mechanism or the fast neutron detection mechanism described above, the same or different amounts of the same activated material are used. A temperature detection mechanism having a structure in which the temperature measurement wire detection medium is fed at the same speed is used.

As the thermal neutron determination wire detection medium and the temperature measurement wire detection medium, Co, which has been shown to be sufficiently usable in the description of thermal neutron flux measurement, is used. Both are detection media containing the same ⁵⁹ Co content. As a thermal neutron collimator made of a material having a large thermal neutron capture cross section, boron nitride having a thickness of 0.3 cm is used. The degree of opening of the window of the thermal neutron collimator is determined depending on the temperature by an opening adjustment mechanism using amber that can withstand high temperatures and a bimetal of bronze.

In this case, the irradiation temperature T and the collimator opening O
Can be defined as O = a · T. Therefore, the thermal neutron flux φr reaching the temperature detection wire detection medium through the thermal neutron collimator
If φ is the thermal neutron flux in front of the thermal neutron collimator, then φr = O · φ = a · T · φ. Here, a is an opening degree coefficient obtained in advance by a calibration experiment.

Accordingly, the amount A activated until the temperature measuring wire is unreeled from the unwinding mechanism and fits in the storage part.
As for r, when F is an activation amount by a unit neutron flux, Ar = φr · F = a · T · φ · F. On the other hand, the amount A _t which is activation to fit in housing portion drawn out from thermal neutrons for wire feeding mechanism, a A _t = φ · F. Therefore, the irradiation temperature T can be obtained by T = (1 / a) · (A _r / A _t ), that is, the ratio of the amount of activation of the temperature measurement wire to the amount of activation of the thermal neutron wire. , The time distribution of the irradiation temperature can be measured using a previously calibrated a.

As described above, carbon as a neutron moderator (heat-resistant temperature = 1000 ° C. or more), boron nitride as a thermal neutron absorber (heat-resistant temperature: 900 ° C. or more), and stainless steel as a detection medium (heat-resistant temperature: 1000 ° C.) ℃ ℃), and a bimetal for high temperature bimetal such as amber and bronze (heat-resistant temperature: 650 ℃)
It becomes possible to measure the time distribution of the irradiation temperature up to.

Further, in place of a bimetal or the like, a device capable of mechanically changing a change in pressure or the like into a change in opening and closing of a collimator can be dealt with in a similar manner.

[0054]

As described above, by using the present invention, it is possible to reduce the thermal neutron flux, the fast neutron flux and the irradiation temperature in the reactor without inserting cables, insertion tubes, etc. into the reactor. It is possible to determine the passage of time. In particular, high irradiation temperature inside the reactor, partition walls between the inside and outside the reactor,
Neutron flux and temperature measurement such as research experiments and material irradiation tests using high-temperature gas furnaces or fast reactors, which have a severe limitation on penetration terms, etc., can be easily performed.

[Brief description of the drawings]

FIG. 1 is a view showing a conventional irradiation ball type neutron measurement system.

FIG. 2 is a diagram showing a reactor neutron flux / temperature detection device of the present invention.

FIG. 3 is a diagram showing a detection block for measuring thermal neutron flux.

FIG. 4 is a graph showing the thermal neutron decay rate with respect to the thickness of boron nitride.

FIG. 5 is a diagram showing a fast neutron flux measurement detection block.

FIG. 6 is a diagram showing a temperature measurement detection block according to the present invention.

FIG. 7 is a view showing a driving mechanism of the present invention.

Claims (20)

[Claims] (A) a wire feeding container surrounded by a thermal neutron absorbing material, a wire detection medium contained in the material and an activated material having a thermal neutron capture cross section, and the detection medium And a wire storage container that is surrounded by a thermal neutron absorbing material that stores the fed wire detection medium, and the wire detection medium passes through the wire neutron irradiation section between the supply container and the storage container. And (b) a detection block including a drive mechanism that draws out and stores the wire detection medium in the wire feeding container at a constant speed using mechanical stress such as a spring or a spring as a power source. Neutron flux / temperature detector in the reactor. 2. The neutron flux in a reactor according to claim 1, wherein a neutron capture cross section of the thermal neutron absorber is equal to or larger than a thermal neutron capture cross section of the activation material.・ Temperature detector. 3. When the feeding of the detection medium of the detecting device according to claim 1 is completed, the wire detecting medium having a length equal to or longer than the neutron irradiation section between the feeding container and the storage container is subjected to thermal neutron absorption. Neutron flux inside the reactor, characterized in that
Temperature detector. 4. The reactor neutron flux / temperature detector according to claim 1, wherein the detection block is separable into a detection mechanism and a drive mechanism. 5. The reactor neutron flux / temperature detector according to claim 1, wherein the detection medium is constantly driven at a constant speed by using a drive mechanism. 6. The neutron flux and temperature detector in a reactor according to claim 1, wherein the detection medium is driven for a predetermined length at predetermined time intervals by using a drive mechanism. 7. The neutron flux / reactor in the reactor according to claim 1, further comprising a device for setting a mainspring of a drive mechanism or a spring as a drive source outside the reactor.
Temperature detector. 8. The detection device according to claim 1, wherein after rewinding the wire detection medium in the detection mechanism separated from the detection block outside the reactor, the radioactivity of the activated wire detection medium is fed out at a constant speed. , Or at regular intervals,
Gamma-ray nuclide analysis is performed by a gamma-ray spectrometer after collimation, and the thermal neutron capture cross section of the activated material in the detection medium, the time when the detection block was irradiated with neutrons in the reactor, and the wire detection medium feeding container and storage Reactor characterized by calculating the time distribution of thermal neutron flux during irradiation based on the residence time in the wire neutron irradiation section between the container and the elapsed time until gamma ray nuclide analysis is performed Inner neutron flux / temperature detector. 9. The detection device according to claim 8, wherein the wire detection medium in the detection mechanism separated from the detection block outside the reactor is once taken out, and set again in a new feeding / storing mechanism. Neutron flux / temperature detector in the reactor that performs the above operations. 10. The detection device according to claim 8, wherein
With respect to the amount of radioactivity for each elapsed time obtained by performing gamma-ray nuclide analysis, the amount of background was obtained based on the amount of radioactivity of the detection medium left in the mechanism for feeding out to the outside in claim 3, and was obtained in claim 8. A neutron flux / temperature detecting device in a nuclear reactor, wherein a thermal neutron flux is corrected and a time distribution of the thermal neutron flux is obtained. 11. The detecting device according to claim 1, wherein a temperature measuring wire having the same content or different content of the same activation material is provided together with a wire detection medium containing the activation material having a thermal neutron capture cross section. A structure in which the detection medium is fed out at the same speed, a thermal neutron collimator made of a material with a large thermal neutron capture cross section is installed in front of this temperature measurement wire detection medium, and the degree of opening of the collimator window is determined by bimetal or the like. A neutron flux and temperature detecting device in a reactor, characterized in that a time distribution of irradiation temperature is obtained by a mechanism which changes according to a temperature of the neutron. 12. The method according to claim 11, wherein the ratio between the amount of radioactivity of the wire detection medium for thermal neutrons irradiated at the same time and the amount of activation radioactivity of the wire detection medium for temperature measurement depends on the irradiation temperature. Neutron flux and temperature detection in a reactor characterized by determining the temperature distribution over time. 13. (a) The fast neutron moderator is set to the outside,
A wire feeding container surrounded by a material with a thermal neutron absorber inside, a wire detection medium containing an activated material with a nuclear reaction cross section for fast neutrons contained in it, and this detection A mechanism for feeding out the medium to the outside, and a storage container surrounded by a material with the high-speed neutron moderator for storing the fed wire detection medium on the outside and the thermal neutron absorbing material on the inside, and the feeding container and the storage container And (b) a mechanical source such as a mainspring or a spring as a power source, and a wire detection medium in a wire feeding container at a constant speed. Extension
A neutron flux / temperature detection cave inside a reactor that includes a detection block composed of a driving mechanism to house it. 14. The neutron flux / temperature detecting device in a nuclear reactor according to claim 2, wherein a time distribution of a fast neutron flux and an irradiation temperature is obtained. 15. The time distribution of irradiation temperature is obtained by using the fast neutron detection mechanism according to claim 13 instead of the thermal neutron flux detection mechanism according to claim 1.
The neutron flux and temperature detector in a nuclear reactor according to the above. 16. The reactor according to claim 1, wherein the same drive mechanism is used to drive the thermal neutron flux measurement or the fast neutron measurement detection mechanism. Inner neutron flux / temperature detector. 17. The detection device according to claim 1, wherein a tape is used instead of a wire as a shape of the detection medium used for thermal neutrons, fast neutrons, or temperature measurement. A neutron flux and temperature detector in a nuclear reactor. 18. The detection device according to claim 1, wherein the activation detection material used for measuring thermal neutrons or fast neutrons has resonances outside thermal neutrons or outside fast neutrons. With a certain neutron capture cross section,
For thermal neutron measurement of claim 1, for temperature measurement of claim 11,
14. A neutron flux / temperature detector in a nuclear reactor, wherein an activation detection material is added as a material surrounding the feeding container and the storage container of the measurement device for high-speed neutron measurement according to claim 13. 19. The detection device according to claim 11, wherein the activation detection material used for thermal neutron measurement includes the activation detection material when the activation detection material has a neutron capture cross section having resonance outside thermal neutrons. Neutron flux and temperature in a reactor characterized by adding the same material as the activation detection material to the collimator material of a thermal neutron collimator made of a material with a large thermal neutron capture cross section in front of the wire detection medium for temperature measurement Detection device. 20. The detection device according to claim 1, wherein two or more materials are used as the activation detection material used for measuring thermal neutrons or fast neutrons. Neutron flux and temperature detector inside the reactor.