KR101339115B1 - Method for measuring delayed neutron in high frequency mode using neutron generator and system thereof - Google Patents

Abstract

The present invention relates to a method and a measurement system for measuring late neutrons in a spent fuel assembly in a high frequency mode using a deuterium-tritium (DT) neutron generator.
A method for measuring late neutrons in a spent nuclear fuel assembly according to the present invention includes: a first step of neutron generating unit generating interrogation neutrons and irradiating the spent nuclear fuel assembly to the spent nuclear fuel assembly; A second step of waiting for irradiated neutrons and prompt neutrons to disappear from the spent fuel assembly to which the irradiated neutrons are irradiated; And a third step of measuring late neutrons generated in the spent fuel assembly after the instantaneous neutron disappearance in the neutron detector.

Description

METHOD FOR MEASURING DELAYED NEUTRON IN HIGH FREQUENCY MODE USING NEUTRON GENERATOR AND SYSTEM THEREOF} Using a neutron generator to measure delayed neutrons in high frequency mode in a spent fuel assembly

The present invention relates to a method for analyzing fissile material in a spent fuel assembly using a neutron generator, and more particularly, in a spent fuel assembly using a deuterium-tritium (DT) neutron generator. The present invention relates to a method and a measuring system for measuring a late neutron generated in a high frequency mode.

In general, in spent fuel material, accurate measurement of the amount of nuclear material is very important to ensure nuclear transparency, and delayed neutrons are used using a neutron generator to determine fissile material in such spent fuel material. neutron, DN) is measured.

Late neutrons represent neutrons that are released during nuclear fission, not with the moment of fission, but with the radioactive decay of fission fragments within a short time after fission.

This measurement technique of late neutrons has been used primarily for new fuel tests and analysis of ²³⁵ U materials in uranium materials over the past few years. The source of late neutrons is the beta decay of fission fragments, and the distribution of fission fragments shortly after fission decreases over time.

Therefore, performing the data acquisition of the late heavy box in the early stage after fission can maximize the efficiency of use of the late neutron. Shuffler was developed at LANL as a representative application for delayed neutron measurement, and the system was developed by PM Rinard, "Shuffler Instrument for the Nondestructive Assay of Fissile Materials," Los Alamos National Laboratory Report, LA-12105 DE91 014281 ( 1991), the measurement accuracy for uranium material is high.

Key points for the measurement of these late neutrons include: (1) investigating neutrons using an external neutron source such as ²⁵² Cf, and (2) interrogation neutrons and prompt neutrons. Wait until is disappeared from the measurement system, start measurement of late neutrons after (3), and (4) repeat the above measurement method until an appropriate measurement error is obtained.

In addition, the conventional delayed neutron measurement system for delayed neutron measurement uses ²⁵² Cf as an active neutron source because the collapse method of ²⁵² Cf is well known, easy to obtain in the market, and high neutron emissivity is high. In particular, delayed neutrons exhibit a lower distribution than immediate neutrons and therefore require high neutron emissivity of the irradiated source to obtain an acceptable error.

However, the ²⁵² Cf radiation source for the measurement of delayed neutron has a disadvantage that the neutron emissivity of the ²⁵² Cf radiation source is reduced and must be replaced with a new one every 5 years. In addition, in a measurement system using a ²⁵² Cf source, a separate mechanical drive system for irradiation control is required, and the ²⁵² Cf source requires a heavy permanent shield.

In addition, in connection with a technique for measuring the concentration of nuclear fission material using a low frequency mode or a deuterium-deuterium (DD) neutron generator, Korean Patent Publication No. 2010-0119194 discloses a nuclear fuel rod using a pulsed DD neutron generator. The present invention relates to a nondestructive testing method for concentration, and includes a sample holder module for accommodating a nuclear material sample, a body portion made of polyethylene, and a neutron detector formed to enclose a sample holder module. However, the DD (deuterium-deuterium) reaction used in the neutron generator of the prior art has a lower yield compared to the deuterium-tritium (DT) reaction, and also causes loss compared to the high frequency mode when the delayed neutron signal is acquired in the low frequency mode.

Therefore, the technical problem of the present invention is to solve the problems described in the prior art as described above, by using a DT (deuterium-tritium) neutron generator as the irradiation neutron source, the maintenance of the neutron generator is relatively complicated, It is an object of the present invention to provide a method and measurement system for measuring late neutrons in a spent fuel assembly with convenient time control and high yield.

The present invention also maximizes signal acquisition by reducing delayed neutron signal loss caused by signal acquisition in low frequency mode and measuring delayed neutrons in high frequency mode to minimize neutron irradiation and delayed neutron signal acquisition cycle of the neutron generator. To provide a method and measurement system for measuring late neutrons in a spent fuel assembly.

However, the object of the present invention is not limited to the above-mentioned objects, and other objects not mentioned can be clearly understood by those skilled in the art from the following description.

In order to achieve the above object, the method for measuring the late neutron in the spent nuclear fuel assembly according to the present invention, the neutron generator generates a first interrogation neutron (interrogation neutron) to irradiate the spent nuclear fuel assembly; A second step of waiting for irradiated neutrons and prompt neutrons to disappear from the spent fuel assembly to which the irradiated neutrons are irradiated; And a third step of measuring late neutrons generated in the spent fuel assembly after the instantaneous neutron disappearance in the neutron detector.

In addition, a system for measuring late neutrons in a spent fuel assembly according to the present invention includes: a neutron generator for generating irradiation neutrons and irradiating the spent fuel assemblies; And at least one neutron detector for detecting delayed neutrons generated after a nuclear fission reaction occurs in the spent nuclear fuel assembly to which the irradiated neutrons are irradiated to generate prompt neutrons. It is done.

According to the method and measurement system for measuring late neutrons in a spent fuel assembly of the present invention, by using a deuterium-tritium (DT) neutron generator as an irradiation neutron source, maintenance of the neutron generator is relatively complicated, and time control is There is an advantage to increase the yield while being convenient.

Further, according to the method and measurement system for measuring late neutrons in a spent nuclear fuel assembly of the present invention, it is possible to reduce late neutron signal loss caused by signal acquisition in a low frequency mode, and to reduce neutron irradiation and delayed neutron signal acquisition cycle of the neutron generator. There is an advantage in maximizing signal acquisition by measuring late neutrons in high frequency mode to make them as small as possible.

1 is a cross-sectional view illustrating a measurement system for measuring late neutrons in a high frequency mode in a spent fuel assembly using a neutron generator according to the present invention.
2 is a perspective view illustrating a measurement system for measuring late neutrons in a high frequency mode in a spent fuel assembly using a neutron generator according to the present invention.
Figure 3 is an exemplary diagram showing the time response of the neutron detector to some spent fuel assemblies with 0-10 Hz irradiation neutron pulses.
4 is an exemplary diagram illustrating time control for irradiation neutron pulses and data acquisition in differential die-away (DDA) and delayed neutron applications.
5 is an exemplary diagram showing delayed neutron distribution changes in the time domain for some other cases of spent fuel assembly.
6 is an exemplary diagram showing a time chart for a repetition period of delayed neutron data acquisition.
FIG. 7 is an illustration showing the time domain variation of the induced fission number of the four major actinide isotopes of interest for two different spent fuel assemblies with an irradiated neutron pulse of 0-10 μs.
FIG. 8 is an illustration showing the cumulative distribution of late neutrons obtained in two different frequency modes for spent fuel assembly with an initial concentration of 5%, 15 GWD / MTU combustion, and 1 year cooling time.
9 is an exemplary diagram showing C1 and C2 coefficients in a late neutron measurement system for 64 spent fuel assemblies.
10 is an exemplary diagram showing delayed neutron counts for 64 simplified spent fuel assemblies.
FIG. 11 is an exemplary diagram showing the delayed neutron count rate of the delayed neutron measurement system for spent fuel assemblies in 64 different cases calculated using the MCNPX code.
FIG. 12 is an illustration showing the effect of ²³⁸ U on late neutron signals on spent fuel assembly in 64 different cases.
FIG. 13 is an exemplary diagram showing delayed neutron counts for spent fuel assemblies in 64 different cases before and after effect correction of ²³⁸ U. FIG.
14 is an illustration showing the signal-to-background ratio of the spent fuel assembly in 52 different cases for late neutron measurements.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Hereinafter, a detailed description of preferred embodiments of the present invention will be given with reference to the accompanying drawings. In the following description of the present invention, detailed description of known functions and configurations incorporated herein will be omitted when it may make the subject matter of the present invention rather unclear. In describing the drawings, similar or identical reference numerals are used for similar or identical components.

Embodiments in accordance with the concepts of the present invention can make various changes and have various forms, so that specific embodiments are illustrated in the drawings and described in detail in this specification or application. However, this is not intended to limit the embodiments in accordance with the concept of the present invention to a particular disclosed form, it should be understood to include all changes, equivalents, and substitutes included in the spirit and scope of the present invention. Also, the terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of the invention.

1 is a cross-sectional view illustrating a measurement system for measuring late neutrons in a high frequency mode in a spent fuel assembly using a neutron generator according to the present invention, and FIG. 2 is a high frequency in a spent fuel assembly using a neutron generator according to the present invention. It is a perspective view which shows the measurement system which measures late neutron in mode.

1 and 2, a measurement system for measuring late neutrons in a high frequency mode in a spent fuel assembly using a neutron generator according to the present invention includes a spent fuel assembly 10, a neutron generator 20 and The neutron detector 30 may be included.

In this embodiment, for example, the spent fuel assembly 10 has a size of 17 × 17, the initial concentration (IE) is in the range of 2, 3, 4, 5%, and the burnability (BU) is in the range of 15, 30. , 45 and 60 GWD / MTU, cooling time (CT) can be set to 1, 5, 20 and 80 years. In addition, the fissile material of the spent fuel assembly 10 may be related to an effective ²³⁹ Pu mass, and in particular, since most of the mass of the spent fuel assembly 10 is ²³⁸ U, it appears from the nuclear fission of ²³⁸ U in the present invention. The influence of delayed neutron signals may also be considered.

The neutron generator 20 generates irradiation neutrons to irradiate the irradiation neutrons generated in the spent fuel assembly 10. The neutron generator 20 includes a DT neutron generator 201 that generates DT (deuterium-tritium) neutrons as irradiation neutrons, a first spectrum control material 202 and a first spectrum regulator disposed outside the DT neutron generator 201. It may include a second strum adjuster 203 disposed outside the spectrum adjuster.

DT neutron generator 201 emits a high energy neutron of 14.1 MeV. The DT reaction (deuterium-tritium reaction) is about 50-100 times higher yield (yiield) than the DD reaction (deuterium-deuterium reaction).

In addition, since the emission energy emitted from the DT neutron generator 201 is very high, at 14.1 MeV, tungsten (W) is disposed outside the DT neutron generator 201 so that the fissile material can be measured in the spent fuel assembly 10. A first spectrum adjuster 202 made of material and a second spectrum adjuster 203 made of beryllium (Be) are provided outside the first spectrum adjuster 202. Thus, the DT neutron source generated in the DT neutron generator 201 is reduced to 1 MeV or less by collisions and reactions with the spectrum modifiers 202 and 203. In particular, it is important to lower the emission energy to 1 MeV or less because the probability of fission of ²³⁸ U is significantly reduced in the energy region of 1 MeV or less.

The neutron detection unit 30 is a delayed neutron generated after the irradiated neutron irradiated from the neutron generator 20 generates a nuclear fission reaction in the spent fuel assembly 10 and promptly neutrons are generated and extinguished. neutron).

Four neutron detectors 30 are provided in the shield 40 made of lead (Pb) material disposed outside the spent nuclear fuel assembly 10.

Each neutron detector 30 may include a ³ He detector 301 that is 3/4 inch in diameter and 2 inches in length to detect late neutrons emitted from the spent fuel assembly 10.

^{The 3} He detector 301 is disposed in a moderator 302 of polyethylene material covered with a Cd liner 303 having a thickness of 0.1 cm. In addition, a Cd liner 303 ′ is disposed between the shield 40 and the neutron generator 20 and the spent fuel assembly 10.

In particular, the Cd liner 303 covering the ³ He detector 301 can measure differential die-away (DDA) neutrons. Thus, the delayed neutron measurement system of the present invention may have a dual or combined application of this DDA neutron and delayed neutron measurement.

In addition, the ³ He detector 301 acquires a signal for the late neutron generated from the spent fuel assembly 10 for 5 to 10 Hz at a high frequency of 100 Hz.

Hereinafter, an example of measuring late neutrons in a high frequency mode in a spent fuel assembly using a neutron generator according to the present invention will be described.

FIG. 3 is an illustration showing the time response of the neutron detector 30 to some spent fuel assemblies 10 with 0-10 μs irradiated neutron pulses, and FIG. 4 in differential die-away (DDA) and delayed neutron applications. An illustration showing time control for irradiation neutron pulses and data acquisition. As shown in the figure, the time interval of DT interrogation neutron pulses is 0 to 10 ms.

Irradiated neutrons disappeared rapidly, and the distribution of the immediate neutrons dropped to the level of late neutrons in the range of 3-5 ms. The overlapping time of the immediate neutron and the late neutron may depend on the fissile material of the spent fuel assembly 10. In addition, through proper time control for data acquisition of irradiated neutron pulses and immediate neutrons and late neutrons, the neutron detector 30 can be used for both DDA and late neutrons.

The time control is as follows. The irradiated neutron pulses are generated for 0 to 10 ms and wait until the neutrons from the neutron generator 20 and the induced nuclear fission of ²³⁸ U are thermal neutronized. Thereafter, data acquisition from 0.2 ms to 1 ms is performed for DDA. Most of the neutrons detected during this interval are high-speed neutrons that are induced by fission of fissile isotopes. Thereafter, it waits for the instant neutron to disappear, and performs data acquisition for the late neutron for 5 kHz to 10 kHz, which may be repeated for 300 seconds at a frequency of 100 Hz. Thus, the total number of cycles may be 30,000.

Usually, the origin of late neutrons can be attributed to the disruption of fission fragments such as ⁸⁷ Br. Species, the collapse of the ⁸⁷ Br ttalhaek Kr ⁸⁷ is released immediately after the collapse of the neutrons ⁸⁷ Br, and are stabilized with ⁸⁶ Kr. This may be an example of various late neutron nuclei after fission has occurred. The late neutron nucleus can be divided into six or eight groups, each group having its own half-life characteristics.

The energy of these late neutrons is generally below 0.5 MeV. Thus, in the present invention, the group decay constant and delayed neutron fraction, in each group for the four main actinide isotopes using a collection of eight groups of parental neutrophils, βi is shown in Table 1 below.

[Table 1]

Here, it can be seen that the contribution to the late neutron signal of ²³⁵ U is greater than the contribution of other nuclides in the same amount. The number of late neutrons decreased with the collapse of the parental nucleus, indicating that the distribution of the late neutron nuclei and late neutrons peaked immediately after fission.

FIG. 5 shows the distribution of delayed neutrons in the time domain for three different cases of spent fuel assembly with neutron irradiation pulses from 0 to 10 Hz. As shown in the figure, the distribution of late neutrons decreases with time.

Conventionally, for such delayed neutron data acquisition, neutrons are irradiated in a short time, the neutron source is removed, and the immediate neutrons are waited for to disappear. After this, neutron irradiation and late neutron measurement for 1 to 2 seconds, delayed neutron measurement is started within 1 second, and this cycle is repeated for a specific time to obtain an appropriate error. In this conventional manner, the mechanical drive system of the ²³⁵ Cf source requires a time interval between irradiation and the start of delayed neutron measurement, but as in the present invention, by employing the neutron generator 301, this time interval is 1000 minutes. It can be reduced at intervals of 1, and it is also possible that an initial measurement for late neutrons is made. This means that high frequency data acquisition for late neutron measurements can be made as compared to conventional methods.

The delayed neutron measurement is performed in an iterative manner to obtain an appropriate error since the delayed neutrons have an inherently low distribution. 6 shows a time chart of repeatedly performing delayed neutron data acquisition. Here, A _i is a late neutron signal obtained in the i _th period. Therefore, the late neutron signal accumulated after the n _th period can be derived by the following equation (1).

[Formula 1]

FIG. 7 shows the time domain variation of the number of induced fissions for the four major actinide isotopes of two different cases of spent fuel assembly with an irradiated neutron pulse of 0-10 μs.

Most of the induced fission here occurs during the time when neutron irradiation begins and the instant neutrons and late neutrons overlap, about 4-5 ㎳. Nuclear fission induced after 5 μs is due to late neutrons. Since the half-life of the delayed neutron hair group exceeds 100 ms, which is significantly longer than 5 ms, it can be assumed that the distribution of the immediate neutron nuclei does not change significantly during the instant neutron decay time interval, 0-5 ms.

The late neutron coefficient for the j _th group mother nucleus of each actinide nuclide obtained during one cycle can be derived by the following equation (2).

[Formula 2]

Where F is the number of induced fissions of each actinide nuclide, ε is the detection efficiency for late neutrons, and β _j and λ _j represent the latent neutron ratio and decay constant for the j _th group parent, respectively. t _DA is the data acquisition time for late neutrons.

In addition, the late neutron coefficient accumulated after n cycles for each actinide nuclide can be derived by the following equation (3).

[Equation 3]

Here, T is the edge time for one cycle.

Equation (3) can be used to calculate the cumulative distribution of late neutrons. In the high frequency mode, the data acquisition system for delayed neutron measurement first has a neutron generation pulse width of 10 ms and waits for 4.99 ms for the irradiation neutron and the immediate neutron to disappear. The data acquisition for late neutrons is then performed for 5-10 Hz, and this cycle is repeated for 300 seconds at a frequency of 100 Hz. Thus, the total number of cycles is 30,000.

Conventional schemes for obtaining late neutron coefficients were performed in low frequency mode with neutron irradiation of 0-0.9 seconds, delay time of 0.1 seconds, and delayed neutron measurements of 1 second. The edge time for this low frequency mode is 2 seconds (eg 0.5 Hz) and the total number of cycles is 150 for 300 seconds.

FIG. 8 shows the cumulative distribution of late neutrons obtained in two different frequency modes for spent fuel assembly with an initial concentration of 5% (IE), 15 GWD / MTU burnup (BU), and 1 year of cooling time. .

This result was obtained using the MCNPX code. As a result, it can be seen that the accumulated distribution of late neutrons is saturated in the range of 150 to 200 seconds, so that the total data acquisition time of 300 seconds is large enough to find the saturated neutron saturation value. The accumulated distribution of late neutrons obtained in the high frequency mode (100 Hz) is 3.25 times higher than the accumulated distribution of late neutrons obtained in the low frequency mode (0.5 Hz).

This indicates that in the high frequency mode, higher values of late neutron ratios can be obtained and the late neutron data acquisition efficiency can be maximized. The mass of each nuclide included in the assembly used for this result is ²³⁸ U-439,634 g, ²³⁵ U-15,684 g, ²³⁹ Pu-1,818 g and ²⁴¹ Pu-146 g.

In addition, the late neutron signal contribution of ²³⁵ U is significantly higher than the other three nuclides because the late neutron ratio is a relatively high value. It should be noted in FIG. 8 that the relatively high amount of ²³⁸ U in the spent fuel assembly does not ignore the contribution of ²³⁸ U to the late neutron signal, and that the contribution of ²³⁸ U is greater than that of ²³⁹ Pu or ²⁴¹ Pu. will be.

This means that the influence of ²³⁸ U on the late neutron coefficient needs to be considered to determine the exact amount of fissile mass contained in the spent fuel assembly by measurement of late neutrons.

The three main fissile nuclides that contribute to late neutron signaling are ²³⁵ U, ²³⁹ Pu, and ²⁴¹ Pu. The total fissile material contained in the spent fuel assembly 10 may be defined as an effective ²³⁹ Pu mass, ²³⁹ Pu _eff .

The concept of ²³⁹ Pue _{ff _ DN} represents the mass of ²³⁹ Pu giving the same reaction obtained from all fissile isotopes in a real sample by using delayed neutron technology, which can then be derived by equation (4).

[Formula 4]

Where ²³⁵ U, ²³⁹ Pu and ²⁴¹ Pu are the mass of each fissile isotope included in the spent fuel assembly 10, and C ₁ and C ₂ are the coefficients for ²³⁵ U and ²⁴¹ Pu, respectively. to be. These coefficients can be defined as the relative signal contributions of the unit masses of the ²³⁵ U and ²⁴¹ Pu isotopes compared to the relative signal contributions of the unit mass of ²³⁹ Pu.

Delayed Neutron Signal Contribution (DNSC), which is a late neutron signal contribution of each isotope, can be derived by the following equation (5).

[Formula 5]

Here, F and ε are obtained using the MCNPX code.

Therefore, the C ₁ and C ₂ coefficients can be derived through the following equations (6) and (7).

[Formula 6]

[Equation 7]

9 shows the C ₁ and C ₂ coefficients of the delayed neutron measurement system for 64 different cases of spent fuel assembly. As shown in FIG. 9, it can be seen that the relative contribution of ²³⁵ U and ²⁴¹ Pu compared to the relative contribution of ²³⁹ Pu is more important for the case of high burnup, low initial concentration, and long cooling time.

To define the concept of the C ₁ and C ₂ coefficients, simplified spent fuel cases are made by removing ²³⁸ U, the three major isotopes and the other remaining nuclides in the spent fuel assembly 10. By doing so, it is possible to exclude the influence of other actinides or fission producing nuclides on late neutron signals.

The delayed neutron counts for 64 cases of the simplified spent fuel assembly are shown in FIG. 10. Here, the intensity of the DT neutron generator 301 is 10 ¹⁰ n / s, the width of the irradiation pulse is 10 ms, the delayed neutron measurement time width is 5 ms (the delay time after the irradiation pulse is 4.99 ms), and the period of one cycle is 10 .

This cycle is repeated for 300 seconds, with a total number of cycles of 30,000. Here, it can be seen that, for several different cases of three fissile isotopes, the comparative amount of ²³⁹ Pu _{eff _ DN} is similar to the late neutron count ratio. This means that the concept of the C ₁ and C ₂ coefficients and the definition of Pu _{eff _ DN} are valid.

<MCNPX Simulation Results>

FIG. 11 shows the delayed neutron counts of the delayed neutron measurement system for 64 different cases of spent fuel assembly calculated using the MCNPX code. The intensity and data acquisition state of the irradiation neutron are the same as in the simplified case.

As shown in FIG. 11, the simplified case includes a 5% initial concentration, a burnup of 45 GWD / MTD, and a 1 year cooling time. The delayed neutron signal of this simplified spent fuel assembly can be significantly reduced by the insertion of other actinides and fission producing nuclides into the spent fuel. This large decrease in late neutron count is due to the neutron absorption effects of these isotopes such as ²⁴⁰ Pu, ²⁴¹ Am, ¹⁵⁵ Gd, and the like.

The effect of the ²³⁸ U nuclide on the late neutron signal can be calculated using Eq. (5).

Results for 64 different cases of spent fuel assembly are shown in FIG. 12. As shown in the figure, these effects are greater for cases of low initial concentration, high burnup, and long cooling time due to the relatively high content of ²³⁸ U of spent nuclear fuel.

The delayed late neutron count is corrected for the effect of ²³⁸ U in FIG. 13. This correction process is necessary to determine the material of the total fission of the spent fuel assembly by delayed neutron measurements. However, spent fuel assemblies are themselves very strong neutron emitters, and most of this background neutron is due to spontaneous fission of the Cm isotope.

Background count rates for the spent fuel assembly case are calculated using the MCNPX code, and the background ratios of 52 cases of the spent fuel assembly for late neutron measurements with irradiation neutrons of 10 ¹⁰ n / s intensity The signal for (background ratio) is shown in FIG.

In FIG. 14, eight cases of 2% IE and four cases of 3% IE were excluded for this analysis because they do not exist, and are shown to exceed background neutrons for several cases of spent fuel assembly through the figure. It can be seen that a strong neutron source is needed.

As described above, in the present invention, a DT (deuterium-tritium) neutron generator is used as an irradiation neutron source, and the delayed neutron is measured in a high frequency mode to maximize signal acquisition.

While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it is to be understood that the invention is not limited to the disclosed embodiments, but, on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims. I will understand. Therefore, the true technical protection scope of the present invention will be defined by the technical spirit of the appended claims.

10: spent fuel assembly 20: neutron generator
30: neutron detection unit 40: shielding material
201: DT (deuterium-tritium) neutron generator
202: first spectrum modifier 203: second spectrum modifier
301: ³ He detector 302: moderator
303, 303 ': Cd Liner

Claims (14)

In a method for measuring late neutrons in a spent fuel assembly,
A first step in which a neutron generator generates interrogation neutrons and irradiates the spent fuel assembly;
A second step in which a nuclear fission reaction occurs in the spent fuel assembly irradiated with the irradiated neutrons to generate prompt neutrons;
A third step of waiting for the irradiation neutron and generated prompt neutron to disappear;
A fourth step of measuring a delayed neutron generated in the spent fuel assembly after the instantaneous neutron disappearance at the neutron detector;
Including;
The neutron detection unit, the delay in the spent nuclear fuel assembly, characterized in that to acquire the data for the late neutron at a frequency that reduces the delayed neutron signal loss, and the neutron irradiation of the neutron generator and the delayed neutron signal acquisition period How to measure neutrons.
The method of claim 1,
The irradiation neutron irradiated in the first step is a deuterium-tritium (DT) neutron, characterized in that the method for measuring late neutrons in spent nuclear fuel assembly.
The method of claim 1,
In the fourth step, the neutron detector is a method of measuring the late neutron in the spent nuclear fuel assembly, characterized in that for obtaining the data for the late neutron at a frequency of 100Hz.
The method of claim 1,
And wherein said pulse of irradiation neutron in said first step is generated for a time of 0 to 10 ms.
The method of claim 3, wherein
Obtaining data for the late neutrons is a method for measuring late neutrons in a spent fuel assembly, characterized in that performed for 5 ~ 10㎳.
The method of claim 5, wherein
Acquiring data for the late neutrons is repeated for 300 seconds.
In a system for measuring late neutrons in a spent fuel assembly,
A neutron generator for generating irradiation neutrons and irradiating the spent fuel assembly; And
At least one neutron detector for detecting delayed neutrons generated after a nuclear fission reaction occurs in the spent nuclear fuel assembly to which the irradiated neutrons are irradiated and prompt neutrons are generated and extinguished;
Lt; / RTI &gt;
The neutron detector is configured to acquire the data for the late neutron at a frequency that reduces the delayed neutron signal loss and decreases the neutron irradiation of the neutron generator and the delayed neutron signal acquisition period, and is disposed outside the spent nuclear fuel assembly. A system for measuring late neutrons in a spent fuel assembly, characterized by being installed in a shield.
The method of claim 7, wherein
The neutron generator,
DT neutron generator for generating deuterium-tritium (DT) neutron as the irradiation neutron;
A first spectrum modulator disposed outside the DT neutron generator;
And a second spectrum modulator disposed outside of the first spectrum modulator.
The method of claim 8,
And the DT neutron generator emits 14.1 MeV of energy. 15. A system for measuring late neutrons in a spent fuel assembly.
The method of claim 9,
The first spectrum control material is made of a tungsten material,
And said second spectrum modulator is comprised of beryllium to reduce the release energy of DT neutrons generated in said DT neutron generator.
The method of claim 7, wherein
And a Cd liner disposed between the spent fuel assembly and the shield.
The method of claim 7, wherein
The neutron detector,
Further comprising a ³ He detector for detecting late neutrons emitted from the spent fuel assembly,
And the ³ He detector is disposed in a polyethylene moderator covered with a Cd liner.
13. The method of claim 12,
And said ³ He detector acquires a signal for late neutrons generated in said spent fuel assembly for 5 Hz to 10 Hz at a frequency of 100 Hz for late neutrons in spent nuclear fuel assembly.
13. The method of claim 12,
The ³ He detector has a diameter of 3/4 inch, a length of 2 inches,
And the Cd liner is 0.1 cm thick.

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