JP5038652B2 - Method for measuring neutron emission rate of neutron emitter and method for confirming neutron characteristics of nuclear fuel - Google Patents

Description

  The present invention relates to a method for measuring a neutron emission rate of a neutron emitter and a method for confirming neutron characteristics of a nuclear fuel. More specifically, the present invention relates to a neutron emission rate measurement method for neutron emitters such as neutron sources such as RI (radioisotope) and accelerators, radioactive materials such as nuclear fuel and radioactive waste, and a neutron characteristic confirmation method for nuclear fuel using the same.

Conventionally, as a method for measuring the neutron intensity (neutron emission rate) of a standard radiation source such as RI (radioisotope), there is a manganese bath method (manganese bath method). In the manganese bath method, a RI neutron source to be measured is placed in a manganese sulfate solution for a certain period of time, ⁵⁶ Mn is generated by a ⁵⁵ Mn neutron absorption reaction, the solution is recovered, and the β Mn of ⁵⁶ Mn contained in the solution is recovered. by measuring the number of γ-rays occurring in decay, by correcting at ^{55 Mn} determined number neutron absorption, ^{55 Mn} and separately neutronics neutron absorption ratio to other substances such as a solution, the neutron intensity Is what you want (
).

Conventionally, the composition and burnup of nuclear fuel are managed based on the fuel composition at the time of manufacture, the combustion time, and the history of the burned nuclear reactor and the like. That is, the history of nuclear fuel is managed by a reactor core management system or the like, and the burnup, composition, etc. of the nuclear fuel are managed based on the history.
English title: “Absolute Measurement of the Neutron Emission Rate with a Manganese Sulphate Bath System”, Author: “PARK H., CHOI KO., LEE JM., LEE KB, HAHN MS (Korea Res. Inst. Standards and Sci. , Daejeon), KRALIK M. (IIR, Prague, CZE) ”, Material Name:“ J Korean Phys Soc ”, JST Material Number:“ T0357A ”, ISSN:“ 0374-4884 ”, Volume Page (Issued Date) ): “Vol.47 No.4 Page.603-609 (October 15, 2005)”, copy table reference: “Copy 4, table 7, reference 11”, document type: “Serial publication (A)”, article Classification: “Original Paper (a1)”, Publication Country: “Korea (KOR)”, Language: “English (EN)”

However, the above-described manganese bath method has the following drawbacks. That is, since the RI neutron source is disposed in the manganese sulfate solution, the manganese sulfate solution adheres to the RI neutron source, and it is necessary to clean the RI neutron source after the measurement. In addition, ⁵⁶ Mn produced by ⁵⁵ Mn neutron absorption reaction is a radioactive substance that β decays with a half-life of 2.58 hours, and although a half-life is relatively short, a large amount of activation solution will be produced. It is necessary to pay attention to and control radiation contamination. In addition, a two-step process of activation of ⁵⁵ Mn and counting of γ rays from ⁵⁶ Mn is required, and it is necessary to move the liquid radioactive substance and time is required for measurement. Further, the neutron absorption rate of ⁵⁵ Mn and manganese sulfate solutions to other substances depends on the energy distribution of the neutron source, and the prediction accuracy largely depends on the accuracy of the nuclear data used in the nuclear calculation.

  In addition, there is a request to confirm whether the composition and burnup of nuclear fuel managed by the reactor core management system is actually accurate.

  An object of the present invention is to provide a method for measuring a neutron emission rate of a neutron emitter capable of measuring a neutron emission rate without contaminating the neutron emitter and a method for confirming neutron characteristics of a nuclear fuel using the neutron emitter. Another object of the present invention is to provide a method for measuring the neutron emission rate of a neutron emitter that is easy to manage radiation, and a method for confirming the neutron characteristics of a nuclear fuel using the method. The present invention also provides a method for measuring the neutron emission rate of a neutron emitter capable of easily measuring the neutron emission rate without moving a liquid radioactive material, and a method for verifying the neutron characteristics of a nuclear fuel using the neutron emission rate. With the goal. Another object of the present invention is to provide a method for measuring the neutron emission rate of a neutron emitter that can eliminate the need to evaluate the energy spectrum of emitted neutrons, and a method for confirming the neutron characteristics of a nuclear fuel using the method. Furthermore, an object of the present invention is to provide a nuclear fuel neutron characteristic verification method that can actually verify whether or not the composition and burnup of the nuclear fuel coincide with those of management.

In order to achieve this object, the method for measuring the neutron emission rate of a neutron emitter according to claim 1 encloses the neutron emitter with a hydrogen-containing material having a thickness sufficient to absorb neutrons emitted from the neutron emitter. A position that causes an absorption reaction of neutrons in a hydrogen-containing substance, and does not affect the neutrons emitted from the neutron emitter by the 2.2 MeV characteristic γ-rays generated by the reaction with hydrogen among the generated γ-rays And the absolute value of the detection efficiency at which the neutron emitted from the neutron emitter generates 2.2 MeV characteristic γ-rays and is counted by the γ-ray detector. The neutron emission rate of the neutron emitter is obtained based on the above.

  Accordingly, neutrons emitted from the neutron emitter are absorbed by the surrounding hydrogen-containing material. At this time, most of the neutron absorption reaction occurs with hydrogen in the hydrogen-containing substance, and as a result, 2.2 MeV characteristic γ rays are generated. The 2.2 MeV characteristic γ rays are counted by a γ ray detector. The gamma ray detector is located at a position that does not affect the neutron emitted from the neutron emitter, that is, a position where the emitted neutron does not reach, in other words, a position outside the thickness necessary for shielding the emitted neutron. Since it is arranged, the γ-ray detector does not affect the neutron absorption reaction of hydrogen.

  Here, the neutron emitted from the neutron emitter reacts with hydrogen to generate 2.2 MeV characteristic γ-ray, and the probability that this characteristic γ-ray is counted by the γ-ray detector is the detection efficiency of the measurement. It can be obtained as an absolute value. Based on the absolute value of the detection efficiency and the count rate of the γ-ray detector, the number of neutrons emitted from the neutron emitter per unit time, that is, the neutron emission rate of the neutron emitter is obtained.

  In addition, the present inventors have newly found that most of the neutrons emitted from the neutron emitter are absorbed by hydrogen in the hydrogen-containing substance, and 2.2 MeV generated by an absorption reaction with hydrogen. It is possible to obtain the neutron emission rate of the neutron emitter with high accuracy based on the characteristic γ ray count rate. In numerical analysis using the continuous energy Monte Carlo method, 99% or more of neutrons emitted from neutron emitters are absorbed by hydrogen-containing materials, specifically, hydrogen in water or hydrocarbon materials, for example, It was confirmed that 2.2 gV characteristic γ rays were generated.

  In this case, as in the method for measuring the neutron emission rate of the neutron emitter according to claim 2, the neutron emitted from the neutron emitter reacts with hydrogen to generate 2.2 MeV characteristic γ-ray, and the characteristic γ-ray The absolute value of the detection efficiency may be obtained on the basis of the probability of reaching the γ-ray detector and the inherent detection efficiency of the γ-ray detector.

  Further, as in the method for measuring the neutron emission rate of the neutron emitter according to claim 3, the relative absorption distribution of the neutrons emitted from the neutron emitter to the hydrogen-containing material, that is, the relative neutrons of hydrogen in the hydrogen-containing material The probability of reaching the γ-ray detector is determined on the basis of the absorptance distribution and the 2.2 MeV characteristic γ-ray attenuation rate in the material existing between the neutron absorption point and the γ-ray detector. Also good.

  The method for measuring the neutron emission rate of a neutron emitter according to claim 4 is to obtain a relative absorptance distribution using a continuous energy Monte Carlo calculation code. Therefore, the relative absorption rate distribution can be obtained with high accuracy by calculation. As the continuous energy Monte Carlo calculation code, for example, MCNP code (Los Alamos National Laboratory, USA), MVP code (Japan, Japan Atomic Energy Research Institute), etc. can be used.

  The method for measuring the neutron emission rate of the neutron emitter according to claim 5 substitutes a nuclear reaction rate similar to the relative absorption rate distribution of hydrogen neutrons as the relative absorption rate distribution. Although it is difficult to directly measure the relative absorptance distribution of hydrogen neutrons, some of the nuclear reaction rates of other elements with neutrons are easy to measure directly and are similar to the relative absorptivity distribution of hydrogen neutrons. is there. By using a nuclear reaction rate that is easy to measure directly instead of the hydrogen neutron relative absorptivity distribution, the relative absorptance distribution can be obtained by measurement.

In this case, as in the method of measuring the neutron emission rate of the neutron emitter according to claim 6, the nuclear reaction rate similar to the absorption rate distribution of hydrogen neutrons is as follows: ⁶ Li neutron absorption rate distribution, ¹⁰ B neutron absorption rate It is preferably one of an absorptance distribution, a ⁶³ Cu neutron absorptance distribution, and a ³ He neutron absorptance distribution.

  According to a seventh aspect of the present invention, there is provided a method for measuring a neutron emission rate of a neutron emitter, wherein a γ-ray detector is moved and characteristic γ-rays are measured at a plurality of locations to obtain a relative absorptance distribution. By measuring 2.2 MeV characteristic γ-rays generated by the neutron absorption reaction of hydrogen while scanning the γ-ray detector, the relative absorptance distribution can be obtained. As a method for obtaining the relative absorptance distribution, for example, an unfolding method such as a back projection method with a filter can be used.

  A method for measuring the neutron emission rate of a neutron emitter according to claim 8 is to obtain a probability of reaching the γ-ray detector using a continuous energy Monte Carlo calculation code. Accordingly, the probability that the emitted neutron generates 2.2 MeV characteristic γ-rays and reaches the γ-ray detector can be calculated with high accuracy. As the continuous energy Monte Carlo calculation code, for example, an MCNP code, an MVP code, or the like can be used.

  In addition, as in the method for measuring a neutron emission rate of a neutron emitter according to claim 9, the hydrogen-containing substance is preferably water or a hydrocarbon substance.

  Further, in the method for measuring a neutron emission rate of a neutron emitter according to claim 10, the neutron emitter is any one of a neutron source, nuclear fuel, and radioactive waste. If the neutron emitted due to the neutron chain reaction occurring inside the neutron emitter and the neutron absorbed inside the neutron emitter can be ignored for the accuracy of measurement, the neutron emitter is considered as the neutron source. On the other hand, if the neutron emitted due to the neutron chain reaction that occurs inside the neutron emitter and the neutron absorbed inside the neutron emitter cannot be ignored for measurement accuracy, the neutron emitter is considered as nuclear fuel or radioactive waste. Think. When the neutron emitter is a neutron source, the neutron emission rate is the neutron intensity. In addition, when the neutron emitter is a nuclear fuel or radioactive waste, the neutron emission rate can be considered to be absorbed by an external hydrogen-containing material. Rate.

  Further, in the method for confirming neutron characteristics of a nuclear fuel according to claim 11, the neutron emitter is a nuclear fuel, and the nuclear fuel is measured using the neutron emission rate measuring method for a neutron emitter according to any one of claims 1 to 9. The neutron emission rate of the nuclear fuel is measured, and the neutron emission rate of the nuclear fuel is calculated on the basis of the burnup of the nuclear fuel predicted for management and the composition based on the burnup, and this value is compared with the measured value of the neutron emission rate. Thus, at least one of the actual burnup of the nuclear fuel and the composition based on the burnup is confirmed.

  Since the nuclear fuel is placed in a sufficiently thick hydrogen-containing material, it can be assumed that all the neutrons released from the nuclear fuel are absorbed by the hydrogen in the hydrogen-containing material. The rate can be regarded as the absorption rate of neutrons in the hydrogen-containing material (the absorption rate of neutrons outside the nuclear fuel). The absorption rate of neutrons in hydrogen-containing materials can be determined with high accuracy by nuclear calculation based on the burnup and composition of nuclear fuel.

The composition of nuclear fuel, that is, the ratio of ²³⁵ U, ²³⁸ U, ²³⁹ Pu, fission products, etc., is managed by a nuclear reactor core management system or the like. Based on the management composition, the absorption rate of neutrons in the hydrogen-containing material is calculated, and the calculated value is regarded as the neutron emission rate and compared with the actual measured value of neutron emission rate. The composition of can be confirmed.

  Similarly, the burnup of nuclear fuel is also managed by the reactor core management system, etc., and the neutron absorption rate in the hydrogen-containing material is calculated based on the burnup on the management, and this calculated value is used as the neutron emission rate. In view of this, it is possible to confirm the burn-up in the management of nuclear fuel by comparing with the actual measured value of neutron emission rate.

The method for measuring a neutron emission rate of a neutron emitter according to claim 1, wherein the neutron emitter is surrounded by a hydrogen-containing material having a thickness sufficient to absorb neutrons emitted from the neutron emitter , Γ-ray detector in which 2.2 MeV characteristic γ-rays generated by the reaction with hydrogen among the generated γ-rays are arranged at positions that do not affect the neutrons emitted from the neutron emitter Of the neutron emitter based on this count rate and the absolute value of the detection efficiency at which the neutrons emitted from the neutron emitter generate 2.2 MeV characteristic γ rays and are counted by the γ ray detector. Since the neutron emission rate is obtained, the neutron can be detected without affecting the emitted neutron, and the neutron emission rate of the neutron emitter can be obtained with high accuracy. That is, in the method of detecting a neutron by arranging a neutron detector in the neutron field, it is necessary to cause the neutron detector to react with the neutron, so that the neutron field is always affected and there is no neutron detector. Measurements cannot be performed on neutron fields. In contrast, in the present invention, the γ-ray detector is disposed at a position sufficiently away from the neutron emitter, and measurement can be performed without affecting the emitted neutrons. It can be obtained with high accuracy.

  Moreover, since characteristic γ rays of 2.2 MeV are promptly emitted from hydrogen that has absorbed neutrons, the neutron emission rate of the neutron emitter can be measured in real time.

  Moreover, in this invention, the neutron emitter to be measured is not polluted like the manganese bathtub method, and the neutron emitter after measurement is not required to be cleaned.

  In addition, since hydrogen atoms that have absorbed neutrons are promptly generated and stabilized by generating 2.2 MeV characteristic γ-rays, activation of hydrogen-containing materials surrounding neutron emitters is at a low level that is not a problem in practice. It is constrained by things, and it is easy to pay attention to and control radiation contamination.

  Furthermore, in the present invention, it is not necessary to evaluate the energy spectrum of neutrons emitted from the neutron emitter. That is, in the manganese bath method, the ratio of neutron absorption to hydrogen, sulfur, and manganese depends on the energy spectrum of neutrons emitted from the neutron source, so when obtaining the total neutron absorption rate from the neutron absorption rate to manganese, It is necessary to evaluate the energy spectrum of neutrons emitted from the neutron. On the other hand, in the present invention, a hydrogen-containing substance is used to absorb neutrons emitted from the neutron emitter, and 99% or more of the emitted neutrons are absorbed by hydrogen in the hydrogen-containing substance. Can be considered absorbed by hydrogen in hydrogen-containing materials. For this reason, it is not necessary to evaluate the energy spectrum of emitted neutrons, and it is easy to obtain the neutron emission rate of the neutron emitter.

  Moreover, the liquid radioactive substance is not moved like the manganese bathtub method, and the measurement of the neutron emission rate is simple.

  In this case, as in the method for measuring the neutron emission rate of the neutron emitter according to claim 2, the neutron emitted from the neutron emitter reacts with hydrogen to generate 2.2 MeV characteristic γ-ray, and the characteristic γ-ray The absolute value of the detection efficiency may be obtained on the basis of the probability of reaching the γ-ray detector and the inherent detection efficiency of the γ-ray detector.

  Further, as in the method for measuring the neutron emission rate of the neutron emitter according to claim 3, the relative absorption distribution of the neutrons emitted from the neutron emitter to the hydrogen-containing material, that is, the relative neutrons of hydrogen in the hydrogen-containing material The probability of reaching the γ-ray detector is determined on the basis of the absorptance distribution and the 2.2 MeV characteristic γ-ray attenuation rate in the material existing between the neutron absorption point and the γ-ray detector. Also good.

  In the method for measuring the neutron emission rate of the neutron emitter according to claim 4, since the relative absorptance distribution is obtained using the continuous energy Monte Carlo calculation code, the relative absorptance distribution is calculated with high accuracy. Can be sought.

  Further, in the method of measuring the neutron emission rate of the neutron emitter according to claim 5, since the nuclear reaction rate similar to the relative absorption rate distribution of hydrogen neutrons is substituted as the relative absorption rate distribution, the relative absorption rate of hydrogen neutrons By using a nuclear reaction rate that can be easily measured directly instead of a distribution, a relative absorptance distribution can be obtained by measurement.

In this case, as in the method of measuring the neutron emission rate of the neutron emitter according to claim 6, the nuclear reaction rate similar to the absorption rate distribution of hydrogen neutrons is as follows: ⁶ Li neutron absorption rate distribution, ¹⁰ B neutron absorption rate It is preferably one of an absorptance distribution, a ⁶³ Cu neutron absorptance distribution, and a ³ He neutron absorptance distribution.

  Further, as in the method for measuring the neutron emission rate of a neutron emitter according to claim 7, the relative absorptance distribution may be obtained by moving the γ-ray detector and measuring characteristic γ-rays at a plurality of locations. .

  In the method for measuring the neutron emission rate of the neutron emitter according to claim 8, since the probability of reaching the above-mentioned γ-ray detector is obtained using a continuous energy Monte Carlo calculation code, the emitted neutron Generates a characteristic γ-ray of 2.2 MeV, and the probability that it reaches the γ-ray detector can be obtained with high accuracy by calculation.

  In addition, as in the method for measuring a neutron emission rate of a neutron emitter according to claim 9, the hydrogen-containing substance is preferably water or a hydrocarbon substance.

  In the neutron emission rate measuring method for a neutron emitter according to claim 10, the neutron emitter is any one of a neutron source, nuclear fuel, and radioactive waste. For measurement accuracy, the measured neutron emission rate is the neutron intensity for neutron sources that can ignore neutrons emitted due to the neutron chain reaction occurring inside the neutron emitter and neutrons absorbed inside the neutron emitter. is there. In addition, when the neutron emitter is a nuclear fuel or radioactive waste, the neutron emission rate can be considered to be absorbed by an external hydrogen-containing material. Rate.

  Furthermore, in the method for confirming neutron characteristics of a nuclear fuel according to claim 11, the neutron emitter is a nuclear fuel, and the nuclear fuel is measured by using the method for measuring a neutron emission rate of a neutron emitter according to any one of claims 1 to 9. The neutron emission rate of the nuclear fuel is measured, and the neutron emission rate of the nuclear fuel is calculated on the basis of the burnup of the nuclear fuel predicted for management and the composition based on the burnup, and this value is compared with the measured value of the neutron emission rate. Confirms at least one of the actual burnup of the nuclear fuel and the composition based on the burnup, so that whether the administrative nuclear fuel composition matches the actual nuclear fuel composition, It can be confirmed whether or not it matches the actual burnup of nuclear fuel. For example, it can be used to verify the composition of nuclear fuel if it is before loading into the reactor, and both the burnup and composition of the nuclear fuel, or both the burnup and composition in the case of spent fuel. Can be used to verify either.

  Hereinafter, the configuration of the present invention will be described in detail based on the best mode shown in the drawings.

FIG. 1 shows the concept of the method for measuring the neutron emission rate of the neutron emitter of the present invention. In this neutron emission rate measurement method, the neutron emitter 1 is surrounded by a hydrogen-containing material 3 having a thickness sufficient to absorb the neutron 2 emitted from the neutron emitter 1, and the neutron 2 is absorbed in the hydrogen-containing material 3. Γ-ray detection in which 2.2 mV characteristic γ-rays 4 generated by the reaction with hydrogen among the γ-rays generated by reaction are arranged at positions that do not affect the neutrons 2 emitted from the neutron emitter 1 The counting rate and the absolute value of the detection efficiency that the neutron 2 emitted from the neutron emitter 1 generates 2.2 MeV characteristic γ-rays 4 and is counted by the γ-ray detector 5. Based on this, the neutron emission rate of the neutron emitter 1 is obtained. In the figure, the neutron 2 and the characteristic γ-ray 4 are shown in the range.

  In this embodiment, the neutron 2 emitted from the neutron emitter 1 reacts with hydrogen to generate 2.2 MeV characteristic γ-ray 4, and the probability that the characteristic γ-ray 4 reaches the γ-ray detector 5; Based on the inherent detection efficiency of the γ-ray detector 5, the absolute value of the detection efficiency is obtained. In addition, the relative absorptance distribution of the neutron 2 emitted from the neutron emitter 1 to the hydrogen-containing substance 3, the neutron absorption location (intersection of the solid line indicated by reference numeral 2 and the dotted line indicated by reference numeral 4 in FIG. 1) and γ The probability of reaching the γ-ray detector 5 is obtained based on the attenuation rate of the characteristic γ-ray 4 of 2.2 MeV in the substance existing between the line detector 5. In this embodiment, the neutron emitter 1 and the γ-ray detector 5 are arranged in the hydrogen-containing substance 3, and the substance existing between the neutron emitter 1 and the γ-ray detector 5 is the hydrogen-containing substance 3. is there.

  The neutron emitter 1 as a measurement object emits neutrons 2 such as a neutron source such as RI (radioisotope), a neutron generation target of an accelerator neutron source, nuclear fuel, radioactive waste, and the like. Here, the neutron emitter 1 when the neutron emitted due to the neutron chain reaction occurring inside the neutron emitter 1 or the neutron absorbed inside the neutron emitter 1 can be ignored for measurement accuracy ( For example, RI, a neutron generation target of an accelerator neutron source, etc.) are used as neutron sources, and a neutron emitter 1 (for example, nuclear fuel, radioactive waste, etc.) when not negligible will be described as nuclear fuel. Note that the neutron source here is not limited to the neutron generation target of RI or accelerator neutron source. For the accuracy of measurement, neutrons emitted due to neutron chain reaction occurring inside and neutrons absorbed inside are used. Any negligible neutron emitter 1 is included. In addition, the nuclear fuel here is not limited to nuclear fuel, radioactive waste, etc. loaded into the nuclear reactor. For the accuracy of measurement, neutrons emitted due to internally generated neutron chain reactions and neutrons absorbed inside Is included if the neutron emitter 1 cannot be ignored. That is, even if the neutron emitter 1 is nuclear fuel, radioactive waste, or the like, in the case of ignoring neutrons emitted due to neutron chain reaction occurring inside and neutrons absorbed inside for the accuracy of measurement The nuclear fuel and radioactive waste may be included in the neutron source. Also, even if the neutron emitter 1 is a RI, a neutron generation target of an accelerator neutron source, etc., neutrons emitted due to neutron chain reactions occurring inside and neutrons absorbed inside are ignored for measurement accuracy If this is not possible, the nuclear fuel may include the RI and the neutron generation target of the accelerator neutron source.

The hydrogen-containing substance 3 is preferably water, a hydrocarbon substance or the like, for example. However, it is not restricted to these, For example, what contains hydrogen, such as alcohol, can be used. In addition, as water, for example, water that normally exists in nature, in other words, water that contains D ₂ O in a small proportion, or light water that does not contain D ₂ O may be used. Moreover, as a hydrocarbon substance, polyethylene, polystyrene, etc. can be used, for example.

  That is, as the hydrogen-containing substance 3, it is possible to use a compound of an element having a small neutron absorption rate and hydrogen. Examples of elements having a small neutron absorption rate include oxygen and carbon, but are not limited thereto. An example of a compound of hydrogen and oxygen is water. As a compound of hydrogen and carbon, for example, there is a hydrocarbon substance. Examples of the compound of hydrogen, oxygen, and carbon include alcohols such as ethanol, ethylene glycol, and glycerin. Further, the hydrogen-containing substance 3 may be made of only hydrogen, that is, hydrogen gas or liquid hydrogen.

  The neutron emitter 1 is disposed in a hydrogen-containing material 3 having a thickness sufficient to absorb the emitted neutron 2 (hereinafter referred to as emitted neutron 2). In the drawing, a sphere 8 indicated by a two-dot chain line conceptually shows a sphere having a radius sufficient to absorb the emitted neutron 2.

  Here, the sufficient thickness means a thickness that does not leak the emitted neutrons 2, that is, a thickness sufficient to shield the emitted neutrons 2. In addition to the thickness that does not completely leak the neutron 2, the thickness includes a thickness that does not significantly affect the measurement result even if some neutron leakage occurs. That is, it means a thickness that can suppress the leakage number of neutrons 2 to such an extent that the accuracy required for measurement can be satisfied. Further, among the neutrons 2 emitted from the neutron emitter 1, for example, a thickness capable of absorbing 99% of the neutrons 2 may be a sufficient thickness. The sufficient thickness is, for example, 30 cm in the case of water, 35 cm in the case of polyethylene as the hydrocarbon material, and more preferably 60 cm in the case of water and 70 cm in the case of polyethylene as the hydrocarbon material. When the neutron emitter 1 is a neutron generation target at the end of the accelerator, the neutron generation target is disposed in the hydrogen-containing material 3. In the present embodiment, water is stored in a pool, and the neutron emitter 1 is disposed in a sufficiently thick water by sinking the neutron emitter 1 in the water.

  The gamma ray detector 5 is a gamma ray detector capable of detecting gamma rays of 2.2 MeV, for example. The γ-ray detector 5 is disposed at a position where the emitted neutron 2 is not affected, that is, a position where the emitted neutron 2 does not reach, in other words, at a position at least sufficiently thick. In the present embodiment, the γ-ray detector 5 is submerged in the pool. However, the γ-ray detector 5 may be arranged outside the pool.

  First, the case where the neutron emitter 1 is a neutron source will be described. In the case of a neutron source, the neutron emission rate to be measured is the neutron intensity because it is not necessary to consider the neutron 2 emitted due to the neutron chain reaction occurring inside or the neutron 2 absorbed inside.

The neutron intensity of the neutron emitter 1 is S (s ⁻¹ ). When the neutron emitter 1 is arranged in a sufficiently thick hydrogen-containing substance 3, it is not necessary to consider multiplication of the emitted neutron 2, and almost all of the emitted neutron 2 (99% or more) is surrounded by surrounding hydrogen. Since it can be absorbed by the contained material 3, the total absorption rate of the neutron 2 in the hydrogen-containing material 3 and the neutron intensity S of the neutron emitter 1 are balanced.

Now, the number of neutrons 2 with direction Ω and energy E passing through the unit area at position r, that is, the neutron flux φ (Ω, E, r), and the absorption rate of neutrons 2 per unit distance into hydrogen. Σ _{a, H} (E, r), the absorption rate of neutron 2 per unit distance into oxygen (when the hydrogen-containing material 3 is water) or carbon (when the hydrogen-containing material 3 is a hydrocarbon) is _expressed as Σ _{a, OorC} Assuming (E, r), the total absorption rate matches the right side of Equation 1.

Of the absorption reaction of the emitted neutron 2 that occurs in the hydrogen-containing substance 3, 99% or more occurs with hydrogen. Therefore, Equation 2 is derived from Equation 1. Then, φ (Ω, E, r) Σ _{a, H} (E, r) is a neutron relative absorptivity distribution (relative absorptivity distribution of neutron 2) by hydrogen, that is, a γ-ray generation rate distribution of 2.2 MeV. .

  Consider arranging the γ-ray detector 5 at a position rd sufficiently away from the neutron emitter 1 and not affecting the neutron chain reaction of the emitted neutron 2. If the attenuation rate of 2.2 MeV characteristic γ-rays 4 in the hydrogen-containing material 3 is τ, the γ-ray flux ψ of the 2.2 MeV characteristic γ-rays 4 on the surface (detection surface) of the γ-ray detector 5 is 3 Here, θ is an angle formed by the normal of the detection surface of the γ-ray detector 5 and the vector rd−r. The vector rd-r is a vector connecting the absorption position of the neutron 2 by hydrogen, that is, the characteristic γ-ray generation position r of 2.2 MeV and the center rd of the γ-ray detector 5.

2.2 Measured by the γ-ray detector 5 when the intrinsic photoelectric efficiency (inherent detection efficiency) of the γ-ray detector 5 with respect to the characteristic γ-ray 4 is ε and the area of the detection surface of the γ-ray detector 5 is s. The count rate D is expressed by Equation 4. The attenuation rate τ is a generally known value, and is about 0.043 cm ⁻¹ for water and about 0.044 cm ⁻¹ for polyethylene which is a hydrocarbon substance. Further, as the intrinsic photoelectric efficiency ε of the γ-ray detector 5 and the area s of the detection surface, the catalog value of the γ-ray detector 5 may be used, or a standard γ-ray source or neutron source with known intensity is used. May be calibrated.

If the count rate D is obtained by measurement and there is information on φ (Ω, E, r) Σ _{a, H} (E, r), the intensity S can be obtained from Equation 2. That is, if there is information on the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} (E, r) of the neutron 2 of hydrogen, the area s, intrinsic photoelectric efficiency ε, angle θ, attenuation rate τ, vector Based on the magnitude of rd-r, the absolute value X1 of the detection efficiency of the characteristic γ ray of 2.2 MeV can be obtained, and the generation rate of 2.2 MeV gamma ray is determined by the absolute value X1 of this detection efficiency and the count rate D. That is, the absolute value (= D / X1) of the neutron absorption rate into hydrogen can be obtained. Then, the intensity S can be obtained from Equation 2 showing the relationship between the absolute value of the neutron absorption rate to hydrogen and the intensity S. The absolute value X1 can be obtained by Equation 5.

<Equation 5>
X1 = sε∫dΩdEdr ³ [φ (Ω, E, r) Σ _{a, H} (E, r) cosθexp (−τ | rr _d |) / {4π | rr _d | ² }]
/ ∫dΩdEdr ³ [φ (Ω, E, r) Σ _{a, H} (E, r)]

Where exp (-τ | rr _d |) / {4π | rr _d | ² } is from a gamma ray source with a relative spatial distribution of φ (Ω, E, r) Σ _{a, H} (E, r) The probability P that the generated characteristic γ-ray 4 reaches the γ-ray detector 5 is preferably corrected by the area s and the angle θ, but the absolute value of the detection efficiency based on the arrival probability P and the intrinsic photoelectric efficiency ε. X1 can be determined.

Further, the relative absorptance distribution of the neutron 2 emitted from the neutron emitter 1 to the hydrogen-containing substance 3, more precisely, the neutron 2 emitted from the neutron emitter 1 that can be regarded as the relative absorptance distribution. The arrival probability P can be obtained based on the relative absorption rate distribution φ (Ω, E, r) Σ _{a, H} (E, r) to hydrogen and the attenuation rate τ. The arrival probability P varies depending on the distance | r−rd | between the characteristic γ-ray generation position r of 2.2 MeV and the center rd of the γ-ray detector 5.

As _a method for obtaining the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} (E, r), for example, there are the following three methods. However, it is not limited to these.

The first method is to obtain a relative absorptance distribution using a continuous energy Monte Carlo calculation code. The governing equation related to the density of neutron 2 is called the Boltzmann equation or the neutron transport method equation. By solving this governing equation, the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} ( It is known that E, r) can be determined. There is a continuous energy Monte Carlo method as a method that can calculate the solution of this equation with the least approximation. As a code for solving the neutron transport equation based on the continuous energy Monte Carlo method, for example, the MCNP code developed by Los Alamos National Laboratory in the United States or the MVP code developed by Japan Atomic Energy Research Institute in Japan can be used. In this embodiment, MCNP-5 of MCNP code, for example, is used to obtain the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} (E, r) of the neutron 2 of hydrogen.

  The second method substitutes a nuclear reaction rate similar to the relative absorption rate distribution of hydrogen neutron 2 as the relative absorption rate distribution. The relative absorptance distribution in the case where the neutron 2 emitted from the neutron emitter 1 is absorbed in the surrounding hydrogen-containing material 3 is difficult to directly measure using a fixed detector.

That is, when the neutron 2 is absorbed by some nucleus, the neutron 2 is converted into one more nucleus. This new nucleus is in an excited state, and it is easy to emit some kind of radiation. ¹⁰ B or the like immediately emits α-particles and becomes nuclei with two fewer neutrons 2 and two protons (α decay). ²³⁵ U and the like cause fission and release fission fragments. ⁶³ Cu or the like is in an excited state for a while to several hours and then emits γ rays. ¹ H and the like generate γ rays simultaneously with neutron absorption.

The issue ^{10 B} and ^{235 U} like charged particles reaction, by measuring the ionization charge by using such ionizing action of the substances according to the charged particles, it is possible to catch a neutron absorption reactions such as ^{10 B} and ^{235 U.} This is a property used as a mechanism of a general radiation detector. The relative absorptance distribution of neutron 2 can be measured by moving ¹⁰ B or ²³⁵ U that can be used as a detector in the ¹⁰ B or ²³⁵ U cluster.

Also, ⁶³ Cu or the like is linearized and placed in a neutron field (field where neutron 2 pours) to be activated. After linear ⁶³ Cu is activated, it is separated from the neutron field and finely separated, and the relative absorption rate distribution of ⁶³ Cu neutrons can be known by measuring the γ-rays generated at each segment (activation method) ).

However, in the case of absorption of neutrons such as ¹ H, no charged particles are generated, so that no ionized charge is generated in situ. Although ionization charges may be generated secondarily by the generation of γ rays, the ionization charge generation point is different from the neutron absorption point, and this makes it impossible to measure the distribution. Moreover, since γ rays generated by the absorption of neutron 2 are generated promptly, the activation method available with ⁶³ Cu or the like cannot be used. Here, it is conceivable to measure γ-rays generated promptly by neutron absorption, but in this case, a method of measuring γ-rays by fixing one γ-ray detector 5 at a certain position, Since the incidence distribution of γ rays is not known, and therefore the detection efficiency of γ rays cannot be obtained, the number of neutron 2 absorption reactions cannot be determined from the number of γ rays measured.

  For this reason, as a result of intensive studies by the present inventors, it was found that there is an element showing a nuclear reaction rate similar to the relative absorptance distribution of neutron 2 of hydrogen. He created the invention of substituting similar nuclear reaction rates. That is, as shown in FIG. 8, the neutron 2 relative absorptivity distribution of a certain element takes a shape very close to the relative absorptivity distribution of neutron 2 of hydrogen. By measuring the relative absorptance distribution of neutron 2, the neutron relative absorptance distribution of hydrogen is obtained.

Similar nuclear reaction rates include, for example, ⁶ Li neutron 2 relative absorption distribution, ¹⁰ B neutron 2 relative absorption distribution, ⁶³ Cu neutron 2 relative absorption distribution, ³ He neutron absorption distribution There is. In particular, the ⁶ Li neutron 2 relative absorptance distribution and the ¹⁰ B neutron 2 relative absorptance distribution are very similar to the hydrogen neutron 2 relative absorptance distribution, and are therefore suitable for substitution. However, it is not limited to the relative absorptance distribution of these neutrons 2, and can be substituted if it can satisfy the accuracy required for measurement even if the degree of similarity is low.

^The measurement of the relative absorptance distribution of neutron 2 of ⁶ Li, the relative absorptance distribution of neutron 2 of ¹⁰ B, the relative absorptance distribution of neutron 2 of ⁶³ Cu, and the absorptance distribution of neutron 2 of ³ He is as follows, for example: It is done.

For example, a Li glass fiber scintillator neutron detector is used to measure the relative absorptance distribution of ⁶ Li neutrons 2. ⁶ Li is placed in the hydrogen-containing substance 3, and when the hydrogen-containing substance 3 is, for example, water, the detector is scanned in water, or when the hydrogen-containing substance 3 is, for example, a solid block of hydrocarbon. A gap is provided in the solid block of hydrocarbon, and the neutron is detected by scanning the gap. By scanning the detector in this way, the relative absorptance distribution of ⁶ Li neutrons 2 can be directly measured by obtaining the neutron count rate at the scanned position.

The measurement of the relative absorptance distribution of ¹⁰ B neutrons 2 uses, for example, a small proportional counter containing boron fluoride gas. For example, ¹⁰ B is placed in water as the hydrogen-containing substance 3, and the small proportional counter is moved in water, and neutrons are measured at each position in the middle of movement. Thus, the relative absorptance distribution of ¹⁰ B neutrons 2 can be directly measured by performing measurement at a plurality of positions while moving the proportional counter.

Further, ⁶³ measuring the relative absorption index of the neutron 2 ^Cu, for example a thin copper wire water, or by providing a gap solid block of hydrocarbons is arranged in the gap, the ⁶⁴ Cu by causing the activation reaction Generate. After a certain time, this copper wire is taken out, the copper wire is cut, and decay gamma rays generated from each piece are measured to obtain the activation rate of the piece of copper wire. And the relative absorption rate distribution of the ⁶³ Cu neutron 2 can be directly measured by making the positional relationship between the cut piece of the copper wire and the field causing the activation reaction correspond.

The ^three measurements of the relative absorption index of the neutron 2 of He is used, for example ³ He gas small proportional counter containing therein a. For example, ³ He is placed in water as the hydrogen-containing substance 3, the small proportional counter is moved in water, and neutrons are measured at each position during the movement. Thus, the relative absorptance distribution of ³ He neutron 2 can be directly measured by measuring at a plurality of positions while moving the proportional counter.

  The third method is to move the γ-ray detector 5 and measure characteristic γ-rays 4 at a plurality of locations to obtain the relative absorptance distribution of hydrogen neutrons 2. For example, as shown in FIG. 2, the γ-ray detector 5 to which the collimator 10 is attached is scanned at a position that does not affect the emitted neutron 2 and the characteristic γ-rays are measured from various positions and angles, thereby comparing the relative neutrons. Measure the absorption distribution. As this method, it is possible to use an unfolding method such as a back projection method with a filter that is used for measuring an X-ray generation rate distribution by γ-ray SPECT, which is a known method. This makes it possible to measure the generation rate distribution of characteristic γ rays 4, that is, the relative absorptance distribution of neutrons 2, using the same principle as that for measuring the X-ray generation rate distribution by γ-ray SPECT.

  Instead of measuring the characteristic γ-ray 4 at a plurality of locations by moving the γ-ray detector 5, the measurement of the characteristic γ-ray 4 at a plurality of locations can be performed by using a plurality of non-moving γ-ray detectors 5. You may make it do.

  In this way, the neutron emitter 1 is submerged in the hydrogen-containing material 3 so that the emitted neutron 2 is absorbed by the hydrogen in the hydrogen-containing material 3 and the generated 2.2 MeV characteristic γ-ray 4 is measured by the γ-ray detector 5. Then, the counting rate D is obtained, and the absolute value X1 of the detection efficiency of the characteristic γ-ray of 2.2 MeV is obtained from the above-mentioned formula 5, and the generation rate of 2.2 MeV gamma rays, that is, the absolute neutron absorption rate to hydrogen is calculated from these. A value (= D / X1) can be obtained. The intensity S of the neutron emitter 1 can be obtained from the relationship of Equation 2. That is, the neutron emission rate when the neutron emitter 1 is a neutron source can be obtained.

For example, a computer can be used to measure the intensity S of the neutron emitter 1. The concept of a measuring device using a computer is shown in FIG. The count rate D measured by the γ-ray detector 5 is input to the computer 12. The computer 12 also receives an absolute value X1 of detection efficiency. The computer 12 calculates the intensity S based on the input count rate D and the absolute value X1 of the detection efficiency, and outputs the result. Instead of inputting the absolute value X1 of the detection efficiency, the arrival probability P, the intrinsic photoelectric efficiency ε, the area s, and the angle θ are input, and the intensity S is calculated by calculating the absolute value X1 of the detection efficiency by the computer 12. May be. Further, instead of inputting the arrival probability P, the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} (E, r) and the attenuation rate τ are input, and the arrival probability P is calculated by the computer 12. The intensity S may be calculated by calculating the absolute value X1 of the detection efficiency. The computer 12 stores a program for calculating the intensity S of the neutron emitter 1 based on the above-described principle, and the intensity based on the input in cooperation with hardware resources such as an arithmetic unit and a storage device constituting the computer 12. S is calculated and output. Data necessary for the calculation is stored in a storage device as a database, for example. Further, the computer 12 may have a communication function, and the above calculation may be performed while communicating with another computer. For example, communication with a server that provides a nuclear calculation service using a continuous energy Monte Carlo calculation code, a database that stores data necessary for calculation, or the like may be performed.

  Next, the case where the neutron emitter 1 is a nuclear fuel will be described. In the case of nuclear fuel, it is necessary to consider the neutron 2 emitted due to the neutron chain reaction occurring inside and the neutron 2 absorbed inside, and the neutron emission rate to be measured is the absorption of neutron 2 outside the nuclear fuel. The rate Avout.

  The region inside the nuclear fuel is Vin, the region outside the nuclear fuel is Vout, the neutron generation rate (a function unique to the composition determined by the burnup) at the position r is S0 (r), and the neutron absorption fission of the nuclear fuel The neutron intensity S due to radioactive decay of nuclear fuel is given by ν, where ν is the number of neutrons generated per unit, Σf (E, r) is the macroscopic fission cross section of nuclear fuel, and Σa (E, r) is the macroscopic neutron absorption cross section of nuclear fuel. Equation 6, Neutral generation rate Schain caused by neutron chain reaction inside nuclear fuel is Equation 7, Neutron absorption rate Avin inside nuclear fuel is Equation 8, Neutron absorption rate Avout outside nuclear fuel is Equation 9

  Further, these relationships are expressed by Equation 10.

  The neutron intensity S, the neutral generation rate Schain, the neutron 2 absorption rate Avin, and the neutron 2 absorption rate Avout are all functions determined by the fuel composition, which is a function of burnup. If there is no information on burnup or composition, none of the neutron intensity S, neutral incidence Schain, neutron 2 absorption rate Avin, and neutron 2 absorption rate Avout can be obtained as absolute values. It was. In the present invention, the absorption rate Avout of the neutron 2 outside the nuclear fuel can be measured.

  Since most of the neutrons 2 emitted from the neutron emitter 1 are absorbed by the hydrogen in the hydrogen-containing material 3, Equation 11 is derived from Equation 9.

If the relative absorptance distribution φΣ _{H, a} , that is, the relative absorptance distribution φ (Ω, E, r) Σa _{, H} (E, r) is obtained by the above method and the count rate D is obtained by measurement, neutron emission The absorption rate Avout of neutrons outside the nuclear fuel of the body 1 can be obtained.

That is, the count rate D obtained by the measurement is expressed by Equation 12, and the absolute value X2 of the detection efficiency when obtaining the absorption rate Avout is expressed by Equation 13. Here, Σ _{a, O} (E, r) is the absorption rate of neutron 2 into oxygen per unit distance.

  That is, the counting rate of the γ-ray detector 5 is D, and the absolute value of the detection efficiency counted by the γ-ray detector 5 when the neutron 2 emitted from the neutron emitter 1 generates the characteristic γ-ray 4 of 2.2 MeV. Is X2, the generation rate of characteristic γ-rays based on the count rate D and the absolute value X2 of the detection efficiency, that is, the neutron absorption rate Avout (= D / X2) outside the nuclear fuel of the neutron emitter 1, that is, Neutron emission rate can be determined.

Where exp (-τ | rr _d |) / {4π | rr _d | ² } is from a gamma ray source with a relative spatial distribution of φ (Ω, E, r) Σ _{a, H} (E, r) The probability P that the generated characteristic γ-ray 4 reaches the γ-ray detector 5 is preferably corrected by the area s and the angle θ. The absolute value of the detection efficiency is based on the arrival probability P and the intrinsic photoelectric efficiency ε. The value X2 can be determined.

Further, the relative absorptance distribution of the neutron 2 emitted from the neutron emitter 1 to the hydrogen-containing substance 3, more precisely, the neutron 2 emitted from the neutron emitter 1 that can be regarded as the relative absorptance distribution. The arrival probability P can be obtained based on the relative absorption rate distribution φ (Ω, E, r) Σ _{a, H} (E, r) to hydrogen and the attenuation rate τ. The arrival probability P varies depending on the distance | r−rd | between the characteristic γ-ray generation position r of 2.2 MeV and the center rd of the γ-ray detector 5.

  In this way, the neutron emitter 1 is submerged in the hydrogen-containing material 3 so that the emitted neutron 2 is absorbed by the hydrogen in the hydrogen-containing material 3 and the generated 2.2 MeV characteristic γ-ray 4 is measured by the γ-ray detector 5. Then, the counting rate D is obtained, and the absolute value X2 of the detection efficiency is obtained from the above formula, and from these, the neutron absorption rate Avout (= D / X2) outside the neutron emitter 1, that is, the neutron emitter 1 is obtained. The neutron emission rate in the case of nuclear fuel can be obtained.

  For the measurement of the neutron absorption rate Avout outside the neutron emitter 1, for example, a computer can be used. The concept of a measuring device using a computer is shown in FIG. The count rate D measured by the γ-ray detector 5 is input to the computer 12.

  The absorption probability Avout may be calculated by inputting the arrival probability P, the intrinsic photoelectric efficiency ε, the area s, and the angle θ as data to the computer 12 and calculating the absolute value X2 of the detection efficiency by the computer 12.

Further, instead of inputting the arrival probability P as data, the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} (E, r) and the attenuation rate τ are input, and the arrival probability P is calculated by the computer 12. Then, the absorption rate Avout may be calculated by calculating the absolute value X2 of the detection efficiency.

  Moreover, you may input the absolute value X2 of detection efficiency as data. That is, when the nuclear fuel composition is known including the spatial distribution, the absolute value X2 of the detection efficiency can be calculated based on the fuel composition, so that even if the calculated absolute value X2 is directly input to the computer 12 good. The computer 12 calculates the absorption rate Avout based on the input count rate D and the absolute value X2 of the detection efficiency, and outputs the result. For example, a continuous energy Monte Carlo calculation code can be used to calculate the absolute value X2 of the detection efficiency when the composition of the nuclear fuel is known including the spatial distribution.

  The computer 12 stores a program for calculating the absorption rate Avout based on the above-described principle, and calculates the absorption rate Avout based on the input in cooperation with hardware resources such as an arithmetic unit and a storage device constituting the computer 12. Calculate and output. Data necessary for the calculation is stored in a storage device as a database, for example. Further, the computer 12 may have a communication function, and the above calculation may be performed while communicating with another computer. For example, communication with a server that provides a nuclear calculation service using a continuous energy Monte Carlo calculation code, a database that stores data necessary for calculation, or the like may be performed.

  In the present invention, since the γ-ray detector 5 is disposed at a position that does not affect the neutron 2 emitted from the neutron emitter 1, the γ-ray detector 5 may affect the emitted neutron 2. Absent. For this reason, the neutron emission rate of the neutron emitter 1 can be obtained with high accuracy. That is, in the method of detecting the neutron 2 by arranging the neutron detector in the neutron field, since the neutron 2 needs to be reacted with the neutron detector, the neutron field is always affected and there is no neutron detector. No measurements can be made on the neutron field in the state. On the other hand, in the present invention, the γ-ray detector 5 is disposed at a position sufficiently away from the neutron emitter 1 and measurement can be performed without affecting the emitted neutron 2. Can be obtained with high accuracy.

  Further, since the characteristic γ-ray 4 of 2.2 MeV is promptly emitted from the hydrogen that has absorbed the neutron 2, the neutron emission rate of the neutron emitter 1 can be measured in real time.

  Moreover, in this invention, the neutron emitter 1 of a measuring object is not polluted like the manganese bathtub method, and washing | cleaning of the neutron emitter 1 after a measurement is unnecessary.

  In addition, since hydrogen atoms that have absorbed neutron 2 are promptly generated and stabilized by generating characteristic γ rays 4 of 2.2 MeV, activation of hydrogen-containing substance 3 surrounding neutron emitter 1 is not problematic in practice. Therefore, attention to radiation contamination and its management are easy.

  Furthermore, in the present invention, it is not necessary to evaluate the energy spectrum of the neutron 2 emitted from the neutron emitter 1. That is, in the manganese bath method, the ratio of neutron absorption to hydrogen, sulfur, and manganese depends on the energy spectrum of neutrons emitted from the neutron source, so when obtaining the total neutron absorption rate from the neutron absorption rate to manganese, It is necessary to evaluate the energy spectrum of neutrons emitted from the neutron. On the other hand, in the present invention, the hydrogen-containing material 3 is used for the absorption of the neutron 2 emitted from the neutron emitter 1, and 99% or more of the emitted neutron 2 is absorbed by the hydrogen in the hydrogen-containing material 3. Therefore, it can be considered that the emitted neutron 2 is absorbed by hydrogen in the hydrogen-containing material 3. For this reason, it is not necessary to evaluate the energy spectrum of the emitted neutron 2, and it is easy to obtain the neutron emission rate of the neutron emitter 1.

  Moreover, the liquid radioactive substance is not moved like the manganese bathtub method, and the measurement of the neutron emission rate is simple.

  Next, a method for confirming neutron characteristics of nuclear fuel will be described.

  This method for confirming the neutron characteristics of nuclear fuel uses the above-described neutron emission rate measurement method for the neutron emitter 1 to measure the neutron emission rate of the nuclear fuel that is the neutron emitter 1 and also predicts the burning of the nuclear fuel that is administratively predicted The neutron emission rate of the nuclear fuel is calculated based on the temperature and the composition determined by the burnup rate, and the calculated burnup value and the measured value of the neutron emission rate are compared to determine the actual burnup of the nuclear fuel and the composition based on the burnup rate. At least one of them is confirmed.

  This method measures the absorption rate of neutrons outside the nuclear fuel determined by the composition determined by the burnup of the nuclear fuel, that is, the neutron emission rate of the nuclear fuel, and confirms the burnup or composition of the nuclear fuel based on the results. To do.

  The number of generations of neutron 2 for nuclear fuel is the generation rate of neutron 2 due to radioactive decay (neutron intensity S) determined for the composition determined by the burnup degree of nuclear fuel, and the neutron 2 due to chain reaction also determined for the composition determined by burnup This is the sum of the incidence Schain. Among the generated neutrons 2, the neutron leakage rate from the nuclear fuel to the surrounding hydrogen-containing substance 3 (the neutron absorption rate Avout outside the nuclear fuel) is also determined by the composition determined by the burnup, and the neutron intensity S and neutron 2 It depends on the incidence Schain. For this reason, the number of neutron leaks to the hydrogen-containing material 3, that is, the absorption rate Avout, in other words, the neutron emission rate, provides strong evidence to support information on the fuel composition or burnup. Therefore, the neutron emission rate can be obtained based on the burnup and composition of nuclear fuel.

  The target nuclear fuel is not limited to a solid state such as spent fuel, but may be a gaseous state or a solution state. Further, the solid state is not limited to the pellet state, and may be a powder state or other states. Moreover, the nuclear fuel before use may be sufficient and a spent nuclear fuel may be sufficient. It is possible to measure the rate (neutron emission rate) of how many neutrons 2 are leaking per unit time around a nuclear fuel or a container containing nuclear fuel in various states.

  The absorption rate Avout of neutrons 2 outside the nuclear fuel, that is, the neutron emission rate, can be obtained with high accuracy by already known nuclear calculations based on the burnup and composition of the nuclear fuel.

  The composition of the nuclear fuel is managed by, for example, a reactor core management system. Based on the management composition, the neutron absorption rate Avout (neutron emission rate) outside the nuclear fuel is calculated, and this calculated value is compared with the measured value of the actual neutron emission rate. The composition can be confirmed.

  Similarly, the burnup of nuclear fuel is also managed by the reactor core management system, etc., and the neutron absorption rate Avout (neutron emission rate) outside the nuclear fuel is calculated based on the burnup on the management, and this calculated value Is compared with the measured value of the actual neutron emission rate, the burn-up in the management of nuclear fuel can be confirmed.

  For example, when the nuclear fuel to be verified is a spent fuel assembly, it can be applied to confirm the burnup of the spent fuel. The fuel assembly used in the light water reactor is stored in a light water pool adjacent to the reactor after being used in the core. The burnup, which is an amount indicating how much energy the fuel used in the light water reactor has released, is evaluated for each fuel assembly by the reactor core management system.

  Once the burnup is determined, the neutron emission rate of the fuel assembly is uniquely determined. By measuring the neutron emission rate of the fuel assembly by the above method and comparing it with the neutron emission rate based on the burnup managed by the core management system, the accuracy of the burnup evaluation value by the core management system is verified. can do. That is, when the measured value and the calculated value based on the core management system are the same or different but within the allowable range, it can be evaluated that the burnup evaluation value by the core management system is accurate. On the other hand, when the measured value and the calculated value based on the core management system are different and the difference exceeds the allowable range, it can be evaluated that the burnup evaluation value by the core management system is not accurate.

  The point where the neutron emission rate is uniquely determined from the burnup is as follows. That is, when the nuclide composition calculation already known based on the burnup (average burnup) managed by the core management system is performed, the composition of the nuclear fuel can be obtained. For this calculation, for example, an ORIGEN code (public calculation code) can be used. The ORIGEN code is a calculation code that can simulate the decrease in the number of fission nuclides due to neutron nuclear reactions in the nuclear reactor, the increase in the number of fission nuclides, and the number of these nuclides further changing due to radioactive decay or neutron nuclear reactions. is there. Based on the calculated composition of nuclear fuel, data on how much neutrons are generated every hour (intensity) from which position (neutron generation position) of the spent fuel is calculated. For this calculation, for example, the ORIGEN code can be used. Furthermore, for the system in which the fuel assembly is submerged in the fuel pool, if the composition information, neutron generation position, and intensity information obtained by the above calculation are input, and neutron transport calculation such as continuous energy Monte Carlo method is performed, The neutron absorption rate in the water surrounding the fuel assembly, that is, the neutron emission rate can be obtained. For this calculation, an MCNP code, an MVP code, or the like can be used.

  It can also be applied to confirm the composition of spent fuel. The composition of spent fuel is evaluated for each fuel assembly by the reactor core management system. As described above, when the fuel composition is determined, the neutron emission rate of the fuel assembly is also uniquely determined. The accuracy of composition evaluation by the core management system is verified by measuring the neutron emission rate of the fuel assembly as described above and comparing it with the neutron emission rate based on the composition managed by the core management system. Can do. In other words, when the measured value and the calculated value based on the core management system are the same or different but within the allowable range, there is a difference between the composition by the core management system and the actual spent fuel composition. It can be evaluated that there is no. On the other hand, if the measured value is different from the calculated value based on the core management system and the difference exceeds the allowable range, the composition between the core management system and the actual spent fuel composition It can be evaluated that there is a habit. Thus, it is effective as means for confirming the presence or absence of wrinkles.

  Further, when the nuclear fuel to be verified is a manufactured MOX fuel assembly, that is, a MOX fuel assembly before use, for example, the following can be applied. That is, when the manufactured MOX fuel assembly is delivered to, for example, an electric power company, specifications regarding the composition of the fuel assembly are reported. When this MOX fuel assembly is submerged in the hydrogen-containing material 3, the number of neutrons 2 leaking into the hydrogen-containing material 3 per unit time (neutron emission rate) is essentially the composition of the MOX fuel assembly. Since it is uniquely determined, the accuracy of the spec can be verified by comparing the neutron emission rate calculated based on the composition on the spec with the actually measured value. That is, if the measured value and the calculated value based on the spec are the same or different but within the allowable range, there should be no difference between the composition according to the spec and the actual composition of the MOX fuel assembly. Can be evaluated. On the other hand, if the measured value is different from the calculated value based on the spec and the difference exceeds the allowable range, there is a flaw between the composition according to the spec and the actual spent fuel composition. Can be evaluated. Thus, it is effective as means for confirming the presence or absence of wrinkles.

  In addition, the accuracy of the composition information of the radioactive material can be evaluated by identifying the neutron emission rate of the neutron emitter 1 for those that emit neutron 2 such as radioactive waste, not limited to nuclear fuel.

The evaluation is performed using, for example, a computer. The concept of the apparatus is shown in FIG. The count rate D measured by the γ-ray detector 5 is input to the computer 12. The computer 12 also receives an absolute value X2 of detection efficiency. The computer 12 calculates an absorption rate Avout (hereinafter referred to as a measured value 14) based on the input count rate D and the absolute value X2 of the detection efficiency. Further, the burnup and composition ratio 13 for management are input to the computer 12. The computer 12 calculates the absorption rate Avout (hereinafter referred to as a calculated value 15) based on the inputted burn-up on management and the composition ratio 13, and evaluates the calculated value 15 by comparing it with the measured value 14. Is output. Instead of inputting the absolute value X2 of the detection efficiency, the arrival probability P, the intrinsic photoelectric efficiency ε, the area s, and the angle θ are input, and the absolute value X2 of the detection efficiency is calculated by the computer 12 to calculate the absorption rate Avout. You may do it. Further, instead of inputting the arrival probability P, the relative absorptance distribution φ (Ω, E, r) Σ _{a, H} (E, r) and the attenuation rate τ are input, and the arrival probability P is calculated by the computer 12. The absorption rate Avout may be calculated by calculating the absolute value X2 of the detection efficiency. Further, instead of inputting the management burnup and composition ratio 13, the absorption rate Avout calculated in advance based on the management burnup and composition ratio 13 is input, and this absorption rate Avout is compared with the measured value 14. You may make it evaluate. The computer 12 is based on the program for calculating the absorption rate Avout based on the above-described principle, the program for calculating the absorption rate Avout based on the burnup and composition ratio 13 on the management, the measured value 14 and the calculated value 15 of the absorption rate Avout. A program or the like for evaluation is stored, and these calculations are performed based on the input in cooperation with hardware resources such as an arithmetic device and a storage device constituting the computer 12. Data necessary for the calculation is stored in a storage device as a database, for example. Further, the computer 12 may have a communication function, and the above calculation may be performed while communicating with another computer. For example, communication may be performed with a server that provides a nuclear calculation service using a continuous energy Monte Carlo calculation code, a server that provides a nuclear calculation service using an ORIGEN code, a database that stores data necessary for calculation, and the like.

  The above-described embodiment is an example of a preferred embodiment of the present invention, but is not limited thereto, and various modifications can be made without departing from the scope of the present invention.

  For example, in the above description, the hydrogen-containing substance 3 is water, and the neutron emitter 1 is arranged in water, but it may be arranged in a hydrocarbon substance instead of water. An example in which the neutron emitter 1 is arranged in a hydrocarbon material is shown in FIG. For example, a hydrocarbon material is filled in a spherical container 6 whose center portion 6a is hollow, and the neutron emitter 1 is placed in the hollow portion of the center portion 6a. The thickness obtained by subtracting the radius of the central portion 6 a from the radius of the spherical container 6 is greater than the thickness sufficient for the hydrogen-containing substance 3 to absorb the emitted neutron 2.

  In this case, the substance existing between the neutron absorption point and the γ-ray detector 5 is the hydrocarbon substance 3 in the spherical container 6 and the air existing outside the spherical container 6, and in some cases, the spherical substance. The air which exists in the center part 6a of the container 6 is also contained. The attenuation rate τ is determined according to the ratio of the thickness of the air and the hydrocarbon material 3.

  When the hydrogen-containing substance 3 is a hydrocarbon substance, the neutron 2 emitted from the neutron emitter 1 is absorbed by the hydrogen in the hydrogen-containing substance 3 in the spherical vessel 6 as in the case where the hydrogen-containing substance 3 is water. Then, the generated 2.2 MeV characteristic γ-ray 4 is measured by the γ-ray detector 5 to obtain the count rate D, and the absolute value X1 or X2 of the detection efficiency is obtained from the above formula, from which the neutron emitter 1 is obtained. The neutron emission rate can be obtained.

  Further, when the neutron emitter 1 itself generates gamma rays, such as nuclear fuel or radioactive waste, for example, a shield 7 is provided as shown in FIG. The γ-ray 9 may be prevented from directly entering the γ-ray detector 5. By preventing the direct incidence of the γ-ray 9 from the γ-ray detector 5, the measurement accuracy of the 2.2MeV characteristic γ-ray characteristic γ-ray 4 can be made higher.

  Alternatively, the arrival probability P may be obtained using a continuous energy Monte Carlo calculation code. That is, by using MCNP code or MVP code, for one neutron emitted from the neutron emitter 1, flight of neutron 2 → absorption reaction → 2.2 MeV characteristic γ ray generation → γ ray detector 5 The arrival probability P may be obtained by calculating the number of arrivals. Then, absolute values X1 and X2 (= P × ε) of detection efficiency may be obtained based on the calculated arrival probability P and the inherent detection efficiency ε of the γ-ray detector.

  Further, in the above neutron characteristic verification method for nuclear fuel, the absorption rate Avout of neutron 2 outside the nuclear fuel is calculated based on the burnup and composition of the nuclear fuel predicted for management, and this calculated value is regarded as the neutron emission rate. Although it was compared with the measured value of the neutron emission rate actually measured, it is not necessarily limited to this method. For example, the measured value of the actually measured neutron emission rate is regarded as the absorption rate Avout, and compared with the calculated value of the absorption rate Avout calculated based on the burnup and composition of the nuclear fuel that is predicted for management. The degree or composition may be confirmed.

In order to confirm the effect of the present invention, calculation analysis was performed. Arrangement in the case of arranging the point-like ^{252 Cf} neutron source as the neutron emitter 1, a point-like ^{252 Cf} neutron source to polyethylene in the center of the sphere of radius 80cm neutron emitter 1 to the center of the sphere of light water radius 80cm Assuming that, the transport of neutrons 2 and the generation of γ rays inside each sphere were calculated. In addition, a point-like γ-ray spectrum detector was placed at a position 76 cm away from the neutron source, and a secondary γ-ray spectrum at the position of the detector was also calculated. For the calculation, a continuous energy Monte Carlo transport calculation code MCNP-5 was used.

  Table 1 shows the calculation results for the neutron 2 absorption site. It is a calculation using the Monte Carlo method, and the calculation result seems to have an error of about 0.3% basically. The “ray source” column indicates that the numerical values in the following column are those when the number of neutrons 2 emitted from the neutron source is 1 (the number of source neutrons is 1). The “Leakage outside the system” column indicates the number of neutrons 2 emitted from the neutron source leaked outside the system. The “number of capture reactions in the medium” column indicates the number of capture reactions generated between the neutrons 2 emitted from the neutron source and the elements in the medium (light water or polyethylene). The “neutron absorption number in hydrogen” column indicates the number of neutrons 2 emitted from the neutron source absorbed by hydrogen in the medium. The “number of γ-rays generated from hydrogen” column indicates the number of neutrons 2 emitted from the neutron source and absorbed by hydrogen in the medium to generate 2.2 MeV characteristic γ-rays 4.

  As is clear from Table 1, the calculated value of the error of about 0.3% may agree with the number of neutrons from the source and the number of γ-rays generated from hydrogen within 1% (0.01). all right. That is, it was newly found that most (99% or more) of the neutrons 2 emitted from the radiation source are absorbed by the hydrogen in the hydrogen-containing substance 3. As a result, it was conceived that the neutron intensity of the radiation source was measured with high accuracy based on the characteristic γ-ray 4 of 2.2 MeV generated by the neutron 2 absorption reaction of hydrogen, and it was confirmed that high-accuracy measurement was possible. .

Relative absorptance distribution of neutron 2 of hydrogen in light water, relative absorptance distribution of neutron 2 of ⁶ Li placed in light water, relative absorptance distribution of neutron 2 of ¹⁰ B, relative of neutron 2 of ⁶³ Cu Absorbance distributions were calculated and the distributions were relatively compared by number. That is, when a point-like ²⁵² Cf neutron source is arranged as the neutron emitter 1 in the center of a light water sphere with a radius of 80 cm, the energy differentiation of the radial neutron flux is calculated, and ⁶ Li, ¹⁰ B, ⁶³ Cu is calculated. The neutron absorption distribution by ⁶ Li, ¹⁰ B, ⁶³ Cu in the radial direction was calculated by multiplying the neutron absorption cross sections for each neutron energy and integrating by energy. For the calculation, a continuous energy Monte Carlo transport calculation code MCNP-5 was used. The result is shown in FIG.

As is clear from FIG. 8, it was found that the relative absorptance distributions of hydrogen, ⁶ Li, ¹⁰ B, and ⁶³ Cu were similar. As a result, it was found that the relative absorptance distribution of hydrogen can be obtained by measuring the relative absorptance distribution of ⁶ Li, ¹⁰ B, and ⁶³ Cu. That is, as the relative absorptance distribution of neutron 2 of hydrogen, the relative absorptivity distribution of neutron 2 of ⁶ Li, similar to the relative absorptivity distribution of neutron 2 of hydrogen, the relative absorptivity distribution of neutron 2 of ¹⁰ B, ⁶³ Cu It was confirmed that the relative absorptance distribution of neutron 2 can be substituted.

In FIG. 8, three graphs of hydrogen, ⁶ Li, and ¹⁰ B substantially overlap when the radial position is within about 3 cm, and when the radial position exceeds about 3 cm, hydrogen, ⁶ Li, ¹⁰ B, Four graphs of ⁶³ Cu almost overlap. FIG. 8 (b) is an enlarged view of the corresponding part of FIG. 8 (a), and FIG. 8 (c) is an enlarged view of the corresponding part of FIG. 8 (a). is there. In FIG. 8A, the substantially overlapping portion is indicated by a solid line. In FIGS. 8B and 8C, the overlapping portions are described in order by line types indicating the respective graphs.

In particular, the relative absorptance distribution of neutron 2 of hydrogen, the relative absorptance distribution of neutron 2 of ⁶ Li, and the relative absorptance distribution of neutron 2 of ¹⁰ B are very similar, and the relative absorptance distribution of neutron 2 of ⁶ Li and It was found that the relative absorptance distribution of neutron 2 of hydrogen was naturally obtained by measuring the relative absorptance distribution of neutron 2 of ¹⁰ B.

The spectrum of γ-rays generated by neutron 2 with a γ-ray detector was examined. Assuming the case where a point-like ²⁵² Cf neutron source is arranged as the neutron emitter 1 at the center of a light water sphere with a radius of 80 cm and a point-like γ-ray spectrum detector is arranged at a position 76 cm away from the ²⁵² Cf neutron source. The spectrum at the detector position of γ rays generated by neutrons emitted from the ²⁵² Cf neutron source was calculated. Here, the influence of γ rays generated from ²⁵² Cf itself is ignored, and the probability that the γ rays generated by neutron 2 being absorbed by light water will be at the detector position is obtained. The characteristic γ-ray flux at the detector position per neutron generated from ²⁵² Cf was calculated. For the calculation, a continuous energy Monte Carlo transport calculation code MCNP-5 was used. The result is shown in FIG.

  As is clear from FIG. 9, it was confirmed that there was a peak (bar shape) at a γ-ray energy of 2.2 MeV. The area of the peak is the probability. As a result, it was found that gamma rays corresponding to 2.2 MeV were clearly obtained. For this reason, it has confirmed that the intensity | strength of the neutron emitter 1 can be measured with high precision by measuring the characteristic γ-ray 4 of 2.2 MeV.

  An experiment was conducted on the hydrogen-containing substance 3 to examine the thickness sufficient to absorb the emitted neutron 2. Specifically, the leakage rate of neutron 2 was calculated when the UOX and MOX spent fuel assemblies for 17 × 17 type PWR having a burnup of 5 to 55 MWD / kg HM were submerged. For the calculation, a continuous energy Monte Carlo transport calculation code MCNP-5 was used.

  As shown in FIG. 10, the neutron flux was calculated with a system in which water (= moderator) as the hydrogen-containing substance 3 was placed around the spent fuel assembly 1 and the outside of the water region was the vacuum boundary 11. . The number of neutrons per unit time (neutron loss rate) absorbed (disappeared) outside the spent fuel assembly 1 was compared with the number of neutrons per unit time (neutron absorption rate) absorbed in the water region.

  The results are shown in Table 2. Table 2 is for a UOX spent fuel assembly for a 17 × 17 PWR with a burnup of 55 MWD / kg HM. In Table 2, the ratio of the “neutron absorption rate in the water region” to the “neutron extinction rate outside the spent fuel assembly” is shown for each “thickness of the water region”. As is clear from Table 2, it was found that when the thickness of the water region is 60 cm, that is, 60 cm away from the surface of the spent fuel assembly, the neutron 2 reaches only about 0.01%. It was also found that the neutron 2 reaches only about 0.1% even when the thickness of the water region is 30 cm. As a result, it was found that when the hydrogen-containing substance 3 is water, the thickness sufficient to absorb the emitted neutron 2 is 30 cm, and preferably 60 cm.

  Next, based on the number of neutrons leaking from the spent nuclear fuel per unit time (neutron leakage rate, that is, neutron emission rate), it was confirmed that information on the burnup of the used nuclear fuel could be confirmed. First, the relationship between neutron leakage rate and burnup of spent fuel was investigated.

  The neutron 2 leakage rate is defined as “neutron absorption rate / neutron generation rate in the peripheral water region” when the spent nuclear fuel assembly is flooded in light water. In the system of FIG. 10, the neutron leakage rate was calculated on the assumption that the thickness of the water (peripheral water region) surrounding the spent nuclear fuel assembly was 60 cm and the nuclide composition was uniform inside the fuel rod. The composition was calculated using the ORIGEN code, and the neutron transport calculation using the continuous energy Monte Carlo transport calculation code MCNP-5.

  The result is shown in FIG. The horizontal axis of the graph in FIG. 11 is the effective multiplication factor keff. The higher the burnup of the spent nuclear fuel assembly, the smaller the nuclear fuel material remaining in the spent nuclear fuel assembly, and the smaller the keff of the system in which the spent nuclear fuel assembly is submerged in light water. Conversely, the lower the burnup of a spent nuclear fuel assembly, the more nuclear fuel material that remains in the spent nuclear fuel assembly, so the keff of the system in which the spent nuclear fuel assembly is submerged in light water increases. Become. That is, the higher the keff, the lower the burnup spent nuclear fuel assembly. As is clear from FIG. 11, the neutron leakage rate hardly changed even when keff changed. As a result, it was found that the neutron leakage rate hardly depends on the burnup.

  Next, the neutron generation rate from the spent nuclear fuel assemblies was calculated for the spent nuclear fuel assemblies having various burnups. The ORIGEN code was used to calculate the number of neutrons generated by composition and radioactive decay, and the continuous energy Monte Carlo transport calculation code MCNP-5 was used to calculate the neutron transport and chain reaction.

  The result is shown in FIG. As is clear from FIG. 12, the neutron generation rate from the spent nuclear fuel assembly differs for each burnup. Since the neutron leakage rate does not depend on the burnup as described above (FIG. 11), it can be said that the neutron leakage rate is almost proportional to the neutron generation rate. For this reason, it was confirmed that the information on the burnup of the fuel could be confirmed by measuring the neutron leakage rate, that is, the neutron emission rate.

For the case where the hydrogen-containing material 3 is polyethylene, a calculation was conducted to examine the thickness sufficient to absorb the emitted neutrons 2. When ²⁵² Cf as the neutron emitter 1 was surrounded by spherical polyethylene, the polyethylene thickness and the absorption rate of the emitted neutron 2 were calculated. For the calculation, continuous energy Monte Carlo transport calculation code MCNP-5 was used.

The results are shown in Table 3. “Thickness” in Table 3 is the radius of the polyethylene sphere. “Absorptance” is a ratio of neutrons 2 absorbed by spherical polyethylene among neutrons 2 emitted from ²⁵² Cf. As is apparent from Table 3, the absorption rate reached 98.44% when the thickness of the polyethylene sphere was 30 cm, and reached 99.63% when the thickness was 40 cm. When the thickness was 70 cm, the absorption rate was 99.99%. As a result, it was found that when the hydrogen-containing material 3 is polyethylene, the thickness sufficient to absorb the emitted neutron 2 is 30 to 40 cm, and preferably 70 cm.

1 shows a first embodiment of a method for measuring a neutron emission rate of a neutron emitter according to the present invention, and is a conceptual diagram when a hydrogen-containing substance is water. FIG. It is a figure which shows the concept which calculates | requires the relative absorption rate distribution of the neutron of hydrogen by moving a gamma-ray detector and measuring a characteristic gamma-ray in multiple places. It is a conceptual diagram of the structure which calculates | requires neutron intensity | strength S. FIG. It is a conceptual diagram of the structure which calculates | requires absorption factor Avout. It is a conceptual diagram of the structure which confirms the neutron characteristic of a nuclear fuel. The 2nd Embodiment of the neutron emission rate measuring method of the neutron emitter of this invention is shown, and it is a conceptual diagram in case a hydrogen containing material is a solid hydrocarbon material. It is a conceptual diagram at the time of providing the shielding body in front of the gamma ray detector which shows 3rd Embodiment of the neutron emission rate measuring method of the neutron emitter of this invention. Relative absorptance distribution of hydrogen neutrons when a point neutron source is placed in light water, ⁶ Li neutron relative absorptance distribution, ¹⁰ B neutron relative absorptance distribution, ⁶³ Cu neutron relative absorptance distribution (A) is a figure which shows the whole, (b) is a figure which expands and shows the corresponding part of (a), (c) is a figure which expands and shows the corresponding part of (a). It is a figure which shows the spectrum in the detector position of the gamma ray generate | occur | produced by a neutron. It is a conceptual diagram which shows the system which sprinkled water (= moderator) as a hydrogen containing substance around the spent fuel assembly. It is a figure which shows the result at the time of calculating a neutron leak rate with the system of FIG. 10, and shows the relationship between an effective multiplication factor keff and a neutron leak rate. It is a figure which shows the result at the time of calculating a neutron generation rate by the system of FIG. 10, and shows the relationship between a burnup and a neutron generation rate.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 Neutron emitter 2 Neutron 3 Hydrogen-containing substance 4 2.2 MeV characteristic γ-ray 5 γ-ray detector

Claims (11)

The neutron emitter is surrounded by a hydrogen-containing material having a thickness sufficient to absorb neutrons emitted from the neutron emitter, causing an absorption reaction of the neutron in the hydrogen-containing material, The characteristic γ-ray of 2.2 MeV generated by the reaction with hydrogen is counted by a γ-ray detector arranged at a position that does not affect the neutron emitted from the neutron emitter, and the counting rate and the neutron emission The neutron emission rate of the neutron emitter is obtained based on the absolute value of the detection efficiency counted by the γ-ray detector by causing the neutron emitted from the body to generate the characteristic γ-ray of 2.2 MeV. A method for measuring the neutron emission rate of a neutron emitter.   The probability that the neutron emitted from the neutron emitter reacts with the hydrogen to generate the characteristic γ-ray of 2.2 MeV and the characteristic γ-ray reaches the γ-ray detector; The method for measuring a neutron emission rate of a neutron emitter according to claim 1, wherein an absolute value of the detection efficiency is obtained based on the intrinsic detection efficiency.   The relative absorptance distribution of the neutrons emitted from the neutron emitter to the hydrogen-containing material, and the characteristic γ-ray of 2.2 MeV in the material existing between the neutron absorption site and the γ-ray detector. The method of measuring a neutron emission rate of a neutron emitter according to claim 2, wherein a probability of reaching the γ-ray detector is obtained based on an attenuation rate.   4. The method of measuring a neutron emission rate of a neutron emitter according to claim 3, wherein the relative absorptance distribution is obtained using a continuous energy Monte Carlo calculation code.   4. The method of measuring a neutron emission rate of a neutron emitter according to claim 3, wherein a nuclear reaction rate similar to the relative absorption rate distribution of hydrogen neutrons is substituted as the relative absorption rate distribution. The nuclear reaction rate similar to the neutron absorption distribution of hydrogen is as follows: ⁶ Li neutron absorption distribution, ¹⁰ B neutron absorption distribution, ⁶³ Cu neutron absorption distribution, ³ He neutron absorption ratio 6. The method for measuring a neutron emission rate of a neutron emitter according to claim 5, wherein the neutron emission rate is one of distributions.   4. The method of measuring a neutron emission rate of a neutron emitter according to claim 3, wherein the γ-ray detector is moved to measure the characteristic γ-rays at a plurality of locations to obtain the relative absorptance distribution.   The method of measuring a neutron emission rate of a neutron emitter according to claim 2, wherein a probability of reaching the γ-ray detector is obtained using a continuous energy Monte Carlo calculation code.   The method for measuring a neutron emission rate of a neutron emitter according to any one of claims 1 to 8, wherein the hydrogen-containing substance is water or a hydrocarbon substance.   The method for measuring a neutron emission rate of a neutron emitter according to any one of claims 1 to 9, wherein the neutron emitter is any one of a neutron source, nuclear fuel, and radioactive waste.   The neutron emitter is a nuclear fuel, and the neutron emission rate of the nuclear fuel is measured using the method for measuring the neutron emission rate of the neutron emitter according to any one of claims 1 to 9, and is predicted for management. The neutron emission rate of the nuclear fuel is calculated based on the burnup of the nuclear fuel and the composition based on the burnup, and the actual burnup and the burnup of the nuclear fuel are compared by comparing the calculated value with the measured value of the neutron emission rate. A method for verifying neutron characteristics of nuclear fuel, characterized in that at least one of the composition based on the above is verified.

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