JP2014174123A - Measurement method of amount of fissile material and measurement device - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a technology that accurately measures an amount of fissile material in a radioactive waste.SOLUTION: A measurement method comprises the steps of: irradiating a radioactive waste with a neutron from a fast neutron source for a predetermined time; measuring the number of fissile neutrons to be detected by a detector disposed at a location outside the radioactive waste; and after irradiating the radioactive waste with the neutron from a neutron generation source, calculating an amount of fissile material in the radioactive waste on the basis of a generation characteristic generated by causing a nuclear reaction of the neutron incident upon the radioactive waste with the fissile material, and peculiar to a measurement device among the number of neutrons detected at a predetermined timing.

Description

  The present invention relates to a technique for measuring the content of fissile material in radioactive solid waste such as a drum can mixed with fissile material such as uranium and plutonium by radiation.

An active neutron method is known as a conventional method for measuring and inspecting the amount of fissile material incorporated in radioactive solid waste by a nondestructive measurement method (for example, Patent Documents 1 to 8, Non-Patent Documents 1 to 3). ).

  In the active neutron method, fast neutrons generated from a neutron source installed in a measurement system are irradiated to radioactive solid waste to be detected, and induce fission reactions with fissionable nuclides in the radioactive solid waste. By measuring the fission neutrons generated as a result of the fission reaction with a neutron detector installed in the measurement system, the fissionable material content of the radioactive solid waste is measured. In particular, the conventional measurement system generates fast neutrons in a pulsed form for a predetermined time and irradiates the fissionable material. Based on the total count of fission neutrons obtained from the pulsed neutron irradiation, The amount of fissile material was derived.

Japanese Patent No. 3845685 JP-A-11-64528 JP 2007-218663 A JP 2009-281878 A Japanese Patent No. 2978103 Japanese Patent No. 2978106 JP-T-2002-541491 JP-A-5-281158

Mitsuo Haruyama et al., Journal of the Atomic Energy Society of Japan, 2001, Vol.43, No.4, p.397-404 Mitsuo Haruyama et al., Japanese Atomic Energy Society Journal, 2004, Vol.3, No.2, p.185-192 Mitsuo Haruyama et al., Japanese Atomic Energy Society Journal, 2007, Vol.6, No.1, p.65-72

  However, even if radioactive solid waste containing the same amount of fissile material, for example, if the neutron absorption characteristics or neutron moderation characteristics of the contents contained in the radioactive solid waste are different, the same amount of neutron irradiation The total count of fission neutrons obtained from may vary. Therefore, it was difficult to accurately measure the amount of fissile material in radioactive solid waste only with the total count of fission neutrons.

  An object of the present invention is to provide a technique for accurately measuring the amount of fissile material in radioactive waste.

In one aspect, the present invention can be exemplified as a method for measuring the amount of fissile material. The method comprises irradiating radioactive waste with a neutron from a fast neutron source for a predetermined time;
Measuring the number of neutrons with a detector located outside the radioactive waste;
Calculating the amount of fissile material in the radioactive waste based on the generation characteristics of the neutrons generated when the neutron incident on the radioactive waste from the neutron source causes a nuclear reaction with the fissile material in the radioactive waste; Have

  According to this method, the amount of fissile material in radioactive waste can be accurately measured.

It is a side view of measuring device 1 concerning an embodiment. 2 is a top view of the measuring device 1. FIG. It is a figure which illustrates the structure of a measurement control PC system. It is a figure which illustrates the count number in a neutron detector. It is a figure which illustrates the procedure of the measurement process of a fissile material amount. It is a figure which illustrates the measurement result in the measuring device of the neutron count number of various radioactive solid waste, and the simulation result which irradiates fast neutron to radioactive waste by neutron transport calculation. It is a figure which illustrates the relationship between the annihilation time obtained by simulation, and a neutron total count number. It is a figure which illustrates the relationship between annihilation time and the neutron total count per unit mass of a fissile material.

  Hereinafter, a method for measuring the amount of fissile material according to an embodiment will be described with reference to the drawings. This measurement method is executed in a measurement device for the amount of fissile material. The configuration of the following embodiment is an exemplification, and the measurement apparatus is not limited to the configuration of the embodiment.

  The measuring apparatus 1 will be described with reference to FIGS. The measuring apparatus 1 performs nondestructive measurement of fissile material in radioactive solid waste using an active neutron method. The measurement apparatus 1 reduces the positional sensitivity difference with respect to the radioactive solid waste in the measurement, and improves the measurement accuracy of the amount of fissile material by considering the neutron absorption and deceleration characteristics of the contents of the radioactive solid waste.

  Specifically, the measuring apparatus 1 derives the amount of fissile material from the annihilation time of the fission neutrons generated from the fissile material and the total count number in order to improve the accuracy of the measurement of the amount of fissile material. As an example of the measurement procedure, the measurement apparatus 1 generates fast neutrons in a pulse form from a neutron generation source and irradiates the solid waste. Then, fission neutrons are generated in pulses from the fissile material contained in the solid waste and reach the neutron detector. Fission neutrons generated at this time decay exponentially with a certain time constant. This time constant is called the die-away time in the active neutron method.

  The extinction time depends on the neutron absorption and moderation characteristics of the substances contained in the radioactive solid waste. For example, the disappearance time is relatively short when the content of the solid waste is a metal system, and the extinction time is longer when the content of the solid waste is concrete or waste (paper or cloth) than that of the metal system. The extinction time is an index indicating the neutron absorption and moderation characteristics of the substances contained in the solid waste. Therefore, the neutron absorption and deceleration characteristics of the substance contained in the radioactive solid waste can be specified by accurately measuring the extinction time in the radioactive solid waste.

  In addition, in response to fast neutrons irradiated in pulses, the total count, which is the total count of fission neutrons generated from the fissile material in the radioactive solid waste and reaching the detector, is detected. Hereinafter, it is simply referred to as a total count number. The total count of fission neutrons is proportional to the amount of fissile material and varies with the neutron absorption and moderation characteristics of the material contained in the solid waste. Therefore, the annihilation of fission neutrons detected in a state containing a known amount of known fissile material by performing a calibration test and changing the characteristic values indicating the neutron absorption and deceleration performance of the material contained in the solid waste The relationship between time and the total count is obtained in advance. For example, the relationship between the annihilation time and the total count of fission neutrons per unit mass of the fissile material is obtained. If this relationship is obtained in advance, it becomes possible to derive an unknown amount of fissile material in solid waste based on this relationship.

  For example, first, the annihilation time of fission neutrons generated from fissile material in solid waste is measured. Next, reference is made to the relationship between the extinction time determined in advance by a calibration test and the total count of fission neutrons per unit mass of fissile material. Then, the total count number of fission neutrons calibrated for the unit mass of the fissile material corresponding to the measured annihilation time is specified. Finally, dividing the total count of fission neutrons from the actually measured fissionable material by the total count of calibrated fission neutrons per unit mass yields the unknown fissionability contained in the solid waste. The amount of substance can be determined.

FIG. 1 and FIG. 2 illustrate the configuration of the measuring apparatus 1. FIG. 1 is a side view of the measuring apparatus 1, and FIG. 2 is a top view. As described above, the measuring apparatus 1 measures the amount of fissile material in radioactive solid waste in a nondestructive manner using the active neutron method. The measuring apparatus 1 includes a table 13 on which a container 6 for storing a fissile material is mounted, a neutron generation source 11 for generating fast neutrons in a pulsed manner and irradiating the fissile material in the container 6 with fast neutrons, A neutron detector 12. Further, the measuring apparatus 1 includes a structure 14 that surrounds the container 6, the table 13, the neutron source 11, and the neutron detector 12, a neutron source controller 2 that controls the neutron source 11, and a neutron detector 12. A neutron pulse signal amplifier 3 for amplifying the detected signal, a neutron pulse signal counter 4 for counting the number of generated neutrons, and a measurement control PC (Personal Computer) 5.

  The container 6 is, for example, a metal columnar box. The container 6 stores a fissile material together with a base material. When measuring the fissile material in the container 6, the container 6 is placed on the table 13. An example of the container 6 is a drum.

  The table 13 desirably has, for example, a turntable that is rotationally driven. That is, it is desirable that the container 6 placed on the table 13 can be rotated around the central axis of the container 6 by rotationally driving the turntable. A turntable is an example of a rotating means.

  The neutron generation source 11 generates fast neutrons in a pulsed manner by a pulsed driving voltage. In the present embodiment, the structure and configuration of the neutron generation source 11 are not limited as long as the neutron generation source 11 generates neutrons in pulses. For example, the neutron generation source 11 extracts deuterium or tritium ions from an ionized gas containing deuterium and tritium ions and collides with a neutron generation target containing deuterium and tritium to generate neutrons. It may be a thing.

  The measurement accuracy of the amount of fissile material by the active neutron method depends on the detection efficiency of fission neutrons. Therefore, the neutron source 11 can generate fast neutrons in a pulsed form with high intensity that can efficiently generate fast neutrons and can sufficiently ensure reproducibility and signal-to-noise ratio in the neutron detector 12. It is desirable to be a thing.

The neutron detector 12 counts the number of fission neutrons when the fission neutrons generated from the container 6 containing the fissile material enter the neutron detector 12. In the present embodiment, the structure and configuration of the neutron detector 12 are not limited. For example, the neutron detector 12 has a structure in which an inert gas such as helium ( ³ He) is introduced between the electrodes. For example, in the neutron detector 12,
What is necessary is just to make a neutron inject into an inert gas and to collect the ion and ion of the inert gas which were ionized to an electrode. The neutron detector 12 may have a configuration in which neutrons are incident on the scintillator, and photons generated by the scintillator are converted into electric signals and counted.

  However, in order to efficiently detect fast neutrons, the neutron detector 12 preferably has a structure proposed by the present inventors in, for example, Japanese Unexamined Patent Application Publication No. 2009-281878. For example, the neutron detector 12 includes a plurality of detectors (A) for detecting neutrons, a thermal neutron absorber (B) such as cadmium surrounding the detector (A), and a thermal neutron absorber (B ) Neutron moderator (C), such as polyethylene, and thermal neutron and epithermal neutron absorber (D) containing boron-containing material surrounding the neutron moderator (C), and thermal neutron and epithermal neutron absorber (D) is a detector bank comprising a structural material surrounding the periphery of (D).

  Thermal neutrons and epithermal neutrons that have entered the detector bank from the space in the structure 14 are absorbed by the thermal neutron and epithermal neutron absorber (D). On the other hand, fast neutrons that have entered the detector bank from the space in the structure 14 are decelerated by the neutron moderator (C) and differentiated into thermal neutrons, epithermal neutrons, or fast neutrons that have not been decelerated. Among these, thermal neutrons changed from fast neutrons are absorbed by the thermal neutron absorber (B). Therefore, the majority of the epithermal neutrons differentiated from the fast neutrons and the fast neutrons that have not been decelerated are detected by the detectors in the detector bank. That is, the neutron detector 12 having the detector bank configuration can selectively detect fast neutrons. Therefore, the neutron detector 12 having the configuration of the detector bank can exclude those that have already become thermal neutrons outside the container 6 and can increase the efficiency with which fast neutrons can be detected.

  For example, the neutron source controller 2 generates a drive voltage in accordance with a pulse waveform control signal, and drives the neutron source 11. Further, the neutron source controller 2 transmits a pulse signal synchronized with a control signal having a pulse waveform for generating a drive voltage to the measurement control PC system 5. The measurement control PC system 5 detects the generation timing of fast neutrons in the neutron generation source 11 based on the transmitted pulse signal. The measurement control PC system 5 may generate a pulse signal and notify the neutron source controller 2 of the pulse signal. In that case, the neutron source controller 2 may generate a pulse waveform drive voltage in accordance with the pulse signal from the measurement control PC system 5 and drive the neutron source 11.

  The neutron pulse signal amplifier 3 amplifies the detection signal of the neutron detector 12 and transmits it to the neutron pulse signal counter 4. The neutron pulse signal counter 4 includes, for example, a comparator that detects a detection signal that can be distinguished from a noise level, and a counter that counts the number of times the comparator is turned on and off. The neutron pulse signal counter 4 notifies the count number counted by the counter to the measurement control PC system 5. However, the measurement control PC system 5 may read the count number from the counter of the neutron pulse signal counter 4.

The measurement control PC system 5 continuously obtains the neutron count number every predetermined time from the timing when the pulsed neutron is generated by the neutron generation source controller 2. For example, the measurement control PC system 5 may be provided with a switch on a circuit for inputting the count number of the neutron pulse signal counter 4. Then, measurement control is performed during a time (ΔT) until a next pulse waveform is generated after a lapse of a predetermined time (T1) from the timing (for example, time 0) when the neutron generation source controller 2 generates pulsed neutrons. The PC system 5 should just turn on a switch and acquire a pulse. However, the measurement control PC system 5 acquires the count from the timing (for example, time 0) when the pulsed neutron is generated by the neutron source controller 2 and controls the neutron source by information processing on a computer. The count number of neutrons after a predetermined time may be obtained from the timing at which pulsed neutrons are generated by the vessel 2.

  The structure 14 includes a fast neutron reflector 14A that is a box-like structure, and a thermal neutron absorber 14B that covers the inner wall of the fast neutron reflector 14A. The fast neutron reflector 14A is, for example, a hexahedron having an internal space. The fast neutron reflector 14A includes at least one of iron, iron alloy, lead, zirconium alloy, and concrete. The thermal neutron absorber 14B is a plate-like material containing at least one of cadmium and boron carbide.

  As the measuring apparatus 1, it is desirable that the position sensitivity difference with respect to the fissile material in the radioactive solid waste is reduced as much as possible. For this reason, in the measuring apparatus 1, the structure 14 surrounding the container 6 to be measured has the thermal neutron absorber 14B lined on the inner wall of the box-shaped fast neutron reflector 14A as the outer frame as described above. It has a structured. Moreover, the space from the inner surface of the thermal neutron absorber 14B lined on the structure 14 to the container 6 is hollow.

  For this reason, the fast neutrons generated by the neutron generation source 11 can pass through the space in the structure 14 and enter the container 6 as it is. Of the fast neutrons generated by the neutron generation source 11, those that do not enter the container 6 penetrate, for example, the thermal neutron absorber 14B and enter the fast neutron reflector 14A. The fast neutrons incident on the fast neutron reflector 14A are reflected and returned to the space from the inner surface of the thermal neutron absorber 14B to the container 6 or decelerated to become thermal neutrons. Then, most of the thermal neutrons remain in the fast neutron reflector 14A or are reflected by the fast neutron reflector 14A and absorbed by the thermal neutron absorber 14B. For this reason, among the fast neutrons generated from the neutron generation source 11, few are slowed down by thermal neutrons and enter the container 6.

  In addition to the fissile material, the container 6 contains a neutron absorbing material such as concrete or polyethylene as a part of the base material. Therefore, the fast neutrons incident on the container 6 are decelerated by the base material in the container 6 and converted into thermal neutrons, causing a fission reaction with the fissile material. As already reported in the Japanese Patent Application Laid-Open Nos. 11-64528 and 2003-90883, fast neutrons are directly incident on the container 6 instead of thermal neutrons, so that the vicinity of the surface of the container 6 and the vicinity of the central axis Thus, the position dependency of the measurement sensitivity of the amount of fissile material is suppressed to about 1/75. The structure 14 in which the thermal neutron absorber 14B is lined on the inner wall of the box-shaped fast neutron reflector 14A shown in FIG. 1 is dependent on the position sensitivity of the measurement sensitivity of the amount of fissile material. It has a structure that can further improve the performance. This is because the thermal neutron absorber 14B absorbs thermal neutrons that cause the position dependence of the measurement sensitivity outside the container 6.

  As illustrated in FIGS. 1 and 2, in the measuring apparatus 1, in order to reduce the position dependency of the neutron detection sensitivity in the radioactive solid waste, the container 6 for storing the radioactive solid waste is provided with a neutron detector 12. It is installed on a table 13 having a turntable with the neutron generation source 11. During measurement of the amount of fission neutrons, it is desirable to rotate the container 6 at a predetermined speed using a turntable. In addition, when the base material of the radioactive solid waste that is the measurement object is a substance having a small self-neutron moderating effect (for example, a metal-based substance), an additional moderator made of polyethylene or the like that is a neutron moderator is connected to the container 6. It is desirable to equip the outside in close proximity. However, the additional moderator made of polyethylene or the like may be enclosed in a state of being separated from the outside of the container 6.

FIG. 3 illustrates the configuration of the measurement control PC system 5. The measurement control PC system 5 includes a CPU 51, a main storage device 52, an external storage device 53, a removable storage device 54, an operation device 55, a display device 56, an input / output interface 57, and a communication interface 58. Have.

The CPU 51 executes a computer program that is executed in the main storage device 52 and provides a function as the measurement control PC system 5. The CPU 51 corresponds to the calculation unit. The main storage device 52 stores a computer program executed by the CPU 51, data processed by the CPU 51, and the like. The external storage device 53 functions as a nonvolatile storage device and stores computer programs, data, and the like stored in the main storage device 52. The external storage device 53 is a hard disk drive, SSD (Solid State Drive), or the like. The removable storage device 54 inputs / outputs a computer program, data, and the like to a removable storage medium. The removable storage medium is a CD (Compact Disc), a DVD (Digital Versatile Disk, Digital Versatile Disc), a USB memory, or the like.

  The operation device 55 is an information input device such as a keyboard, and a pointing device such as a mouse and a touch panel. The display device 56 is, for example, a liquid crystal display, an EL (electroluminescence) display, or the like.

  The input / output interface 57 is, for example, a data input / output interface between the neutron generation controller 2 and the neutron pulse signal counter 4. The input / output interface 57 receives a pulse signal synchronized with the drive voltage pulse from the neutron generation controller 2 to the neutron generation source 11. Further, the input / output interface 57 acquires the count number of neutrons detected by the neutron detector 12 from the neutron pulse signal counter 4.

  The input / output interface 57 obtains the neutron count after a predetermined time from the timing when the pulsed neutron is generated by the neutron source controller 2 among the neutron count detected by the neutron detector 12. You may have the switch for doing. However, the CPU 51 obtains the time (time 0) of the timing of the pulse signal synchronized with the drive voltage pulse from the neutron generation controller 2 to the neutron generation source 11 and the count number for each predetermined measurement interval from time 0. May be. Then, the CPU 51 extracts the neutron count after a predetermined time from the timing when the pulsed neutron is generated by the neutron source controller 2 from the neutron count detected by the neutron detector 12. May be.

  The communication interface 58 is, for example, a LAN (Local Area Network) card. The communication interface 58 connects the measurement control PC system 5 to a communication device on the network.

  In FIG. 4, the count number in the neutron detector 12 which CPU51 acquires is illustrated. This count number is an example in which the neutron count number is extracted from the timing at which pulsed neutrons are generated by the neutron source controller 2 out of the neutron count numbers detected by the neutron detector 12. is there. That is, the neutron count number illustrated on the vertical axis in FIG. 2 is synchronized with the neutron irradiation start time (for example, time 0) from the neutron generation source 11 in FIG. (T1) represents the neutron count per unit time detected by the neutron detector 12 with a delay.

  As described above, the thermal neutron component induced in the measurement apparatus 1 is almost removed due to the configuration of the measurement apparatus 1. Therefore, the measured neutron components are a fast neutron component (L1) from the neutron generation source and a fission neutron component (L2) from the fissile material as shown in FIG.

The fast neutron component (L1) from the neutron generation source 11 is generated by the neutron generation source 11 operating at high speed in a pulsed manner in an extremely short time of 10E-5 seconds or less to generate fast neutrons. As shown in FIG. 4, the fast neutron component (L1) decelerates at a relatively early time and changes to thermal neutrons, and thus shows a rapid decay.

  On the other hand, the fission neutron component from the fissile material in the container 6 is caused by the fission in the container 6. That is, the fast neutrons pulsed on the container 6 from the neutron generation source 11 penetrate into the radioactive solid waste in the container 6, are decelerated by the base material in the container 6, and change into thermal neutrons to be radioactive. Present in solid waste. The thermal neutrons in the vessel 6 are attenuated while inducing a fission reaction with the fissile material and releasing fast neutrons (hereinafter, fission neutrons).

  The amount of fission neutrons emitted from the fissile material is also attenuated in accordance with the attenuation of the amount of thermal neutrons present in the radioactive solid waste in the container 6. This decay time (T) is called the extinction time (T). In other words, the fission neutron annihilation time (T) corresponds to the time at which the fission neutron count is 1 / e. The annihilation time (T) of fission neutrons depends on the neutron absorption and moderation characteristics of the contents contained in the radioactive solid waste. However, the annihilation time (T) of fission neutrons is longer than the decay time of the fast neutron component (L1). Therefore, as shown in FIG. 4, the count number of the fission neutron component (L2) can be distinguished from the fast neutron component (L1) from the neutron generation source 11. The measuring device 1 separates the neutron count number counted by the neutron pulse signal counter 4 and acquired through the input / output interface 57 into a fission neutron component (L2) and a fast neutron component (L1) from the neutron generation source 11. To do. For example, the CPU 51 of the measuring apparatus 1 may calculate the straight line of the fast neutron component (L1) by the least square method or the like in the region where the fast neutron component (L1) from the neutron generation source 11 is dominant. The CPU 51 of the measuring apparatus 1 may calculate a straight line of the fission neutron component (L2) by the least square method or the like in a region where the fission neutron component (L2) is dominant.

  Moreover, the neutron total count number (S) illustrated by the mesh of FIG. 4 can be acquired by integrating the neutron count number per unit time detected by the neutron detector 12. The total neutron count (S) is proportional to the amount of fissile material in the radioactive solid waste since there is no positional sensitivity difference.

  With the above procedure, the CPU 51 of the measuring apparatus 1 calculates the annihilation time (T) of the fission neutron component (L2) and its total neutron count (S), analyzes it, and analyzes the fissionability in the radioactive solid waste. Specify the amount of substance.

  In measuring the amount of fissile material by the measuring device 1, a preliminary calibration test is performed. For example, the relationship between the extinction time (T) and the total neutron count (S) in FIG. 2 obtained by the measuring apparatus 1 in a state where a known fissile material is contained in radioactive solid waste is obtained in advance. . In the calibration test, the relationship between the extinction time (T) and the total neutron count (S) is determined in advance for a plurality of radioactive solid wastes by changing the density and type of the base material in the container 6.

  As illustrated in the following example, the relationship between the annihilation time (T) of fission neutrons and the total neutron count (S) is less dependent on the type and density of the base material contained in the radioactive solid waste. . That is, the relationship between the annihilation time (T) of fission neutrons and the total neutron count (S) can be said to be a unique correlation for each measuring device 1. Therefore, the correlation between the annihilation time (T) of the fission neutrons and the total neutron count (S) in the measuring device 1 from a plurality of radioactive solid wastes having different types and densities of the base material is determined by a prior calibration test. Obtained from data at multiple measurement points. Then, this correlation is made into a database using a table or the like. Alternatively, a correlation equation may be obtained from the relationship between the annihilation time (T) of fission neutrons and the total neutron count (S).

Through the calibration test as described above, the CPU 51 of the measuring apparatus 1 stores the database indicating the correlation, the coefficient of the correlation equation, or the like in the main storage device 52, the external storage device 53, or the like.
As a result, the CPU 51 can refer to the total number of neutrons (S) in the fissile material per unit mass corresponding to the annihilation time (T) in a predetermined range as a reference value.

  FIG. 5 exemplifies the procedure of the measurement processing of the amount of fissile material in the container 6 by the measurement control PC system 5. The procedure of FIG. 4 is executed by the CPU 51 according to a computer program, for example. In this process, first, the CPU 51 acquires the count number of fissile neutrons (S1). As described above, the CPU 51 may acquire the count of neutrons after a predetermined time has elapsed from the timing at which pulsed neutrons are generated by the neutron generation source controller 2. However, the CPU 51 may acquire the neutron count number in the region of the straight line L2 illustrated in FIG. 4 from the neutron count number detected by the neutron pulse signal counter 4. For example, the CPU 51 may obtain the straight lines L1 and L2 from the neutron count number and obtain the neutron count number in the region on the right side from the intersection of the two straight lines. The CPU 51 that executes the process of S1 is an example of a unit that calculates the number of generations of fission neutrons.

  Next, the CPU 51 measures (calculates) the extinction time T from the fission neutron component (L2) illustrated in FIG. 4 (S2). The CPU 51 that executes the process of S2 is an example of a unit that calculates the disappearance time.

Note that a database or correlation formula of the total count number (Nu) of fission neutrons generated from the fissile material at the unit mass with respect to the annihilation time T is set in the main storage device 52, the external storage device 53, etc. (PRE). The main storage device 52, the external storage device 53, and the like that store the database or the correlation equation are examples of means for storing the correlation.

Therefore, the CPU 51 derives the total count number (Nu) of fission neutrons generated from the fissile material per unit mass in the measuring device 1 corresponding to the extinction time T (S3). The CPU 51 that executes the process of S3 is an example of a means for obtaining a total calibration count.

  On the other hand, the CPU 51 measures the total fission neutron count number (Nt) (S4). The fission neutron total count number (Nt) may be obtained by a hardware counter in the measuring apparatus 1 or may be added to the fission neutron count number acquired by the CPU 51. For example, the CPU 51 performs a time (ΔT) until a next pulse waveform is generated after a predetermined time (T1) has elapsed from the timing (for example, time 0) when the pulsed neutron is generated by the neutron generation source controller 2. The count number may be added. This is an example of a means for calculating the total count number by the CPU 51 that executes the process of S4.

Then, the CPU 51 divides the total number of fission neutrons (Nt) by the total number of fission neutrons (Nu) generated from the fissile material at the unit mass (1 g) (S5). This is an example of means for deriving the amount of fissile material by the CPU 51 that executes the process of S5. Moreover, it is an example of a means which CPU51 which performs the process of S5 divides. With the above procedure, the CPU 51 calculates the mass Mf of the fissile material contained in the radioactive solid waste, and outputs it to the display device 56, for example.

(Equation 1)
Mass of fissile material contained in radioactive solid waste Mf = Nt / Nu;
Here, the process of automatically calculating the mass Mf of the fissile material contained in the radioactive solid waste by the CPU 51 of the measurement control PC system 5 is exemplified. However, for example, the measurement control PC system 5 may output the count number illustrated in FIG. That is, the measurement control PC system 5 may be a simple measuring device that outputs a signal from the neutron pulse signal counter 4. That is, the procedure after the detection value of the neutron is obtained from the radioactive waste in the container 6 by the configuration of FIGS. 1 and 2 may be any procedure. For example, a part of the procedure illustrated in FIG. 5 may be executed manually.

  An example of the measuring apparatus 1 will be described with reference to FIGS. 4 and 6-8. Using the measurement apparatus 1, the relationship between the amount of fissile material in the radioactive solid waste, the annihilation time illustrated by T in FIG. 4, and the total neutron count of the fast neutron components illustrated by S in FIG. Asked for. As a result, it was found that there was a clear correlation between the amount of fissile material in radioactive solid waste, the extinction time, and the total neutron count.

  The total neutron count (S) in FIG. 4 is proportional to the amount of fissile material in the radioactive solid waste. Moreover, it turned out that the extinction time (T) of FIG. 4 is dependent on the characteristic value (for example, material density etc.) which shows the neutron absorption and deceleration of the substance which is the content contained in radioactive solid waste.

  Further, FIG. 6 shows the actual measurement results when the iron material is used for the contents of the 200 liter drum in the measuring device 1 for the neutron count number of various radioactive solid wastes, and the radioactive wastes are calculated by the neutron transport calculation. And simulation results of irradiation with fast neutrons. In FIG. 6, the horizontal axis is the time from the generation of neutrons at the neutron generation source 11, and the vertical axis is the neutron count number detected by the neutron detector 12.

  The model structure of the radioactive solid waste and measurement device used in the simulation is as follows. As the structure 4 in FIG. 1, an iron reflector having a thickness of 20 cm (14A in FIG. 1) was provided, and a polyethylene reflector having a thickness of 10 cm was provided around the iron reflector. Moreover, as the neutron detector 12, five detector banks (including 30 detectors in total) in which six detectors are arranged on a straight line are installed. A neutron source 11 was placed at a position facing the detector bank with the radioactive solid waste in between.

  As a container 6, a 200-liter drum was filled with the following three types of base materials, and a simulation was performed.

(Base material 1) Wes-based test body A 200-liter drum can was filled with waste as a simulated base material in a range of 0.1 to 0.6 g / cubic centimeter, and 72 g of uranium 235 (U-235) was set. The simulation was performed under the conditions.

(Base material 2) Metal-based specimen A 200-liter drum can was filled with iron as a simulated base material in a range of 0.1 to 2.0 g / cubic centimeter, and 72 g of uranium 235 (U-235) was set in the center. The simulation was performed under the condition that a polyethylene moderator was set to a thickness of 20 mm around the drum.

(Base material 3) Concrete specimen Specimen was filled in a 200-liter drum can as a simulated base material at a filling density of 0.5 to 2.2 g / cubic centimeter, and 72 g of uranium 235 (U-235) was set in the center. The simulation was performed under the condition that a polyethylene moderator was set to a thickness of 20 mm around the drum.

  As shown in FIG. 6, it was confirmed that the measurement test result and the simulation result were in good agreement with the time change of the neutron count number, and the test result could be reproduced well by the simulation.

In addition, the correlation between the annihilation time and the total neutron count of the fast neutron component was also determined by simulations with varying material densities in areas not tested. As a result, a characteristic value indicating the neutron absorption and deceleration of a wider range of materials than the test, for example, the relationship between the annihilation time obtained so far and the total neutron count of fast neutron components (e.g., different material types) It was also confirmed that it can be applied in the area of the material density). An embodiment of a method for deriving the amount of fissile material in radioactive solid waste will be described focusing on the simulation results.

As a fissile material, net 72g of U-235 is installed in the center of drum-type radioactive solid waste (uniformly distributed in 200cc space), and the content of waste is metal (iron), waste (cloth, etc.)
And concrete rubble etc. were individually installed in a 200 liter drum, and the calculation was performed by changing the material density of the contents.

The irradiation conditions of the fast neutron source 11 in FIG. 1 were such that 10E + 6 fast neutrons with an energy of 14 MeV per pulse were generated at 100 pps and operated for 100 seconds. Here, pps is the number of pulses per second. Then, the extinction time (T in FIG. 4) and the total neutron count (S in FIG. 4) of the fast neutron component (L2 in FIG. 4) measured for 100 seconds were calculated. FIG. 7 illustrates the relationship between the annihilation time and the total neutron count obtained as a result. From Fig. 7, the relationship between the extinction time and the total neutron count is distributed in almost the same straight line on the graph even in the case of concrete systems containing contents such as metal (iron), waste (cloth), concrete rubble, etc. Turned out to be the result. That is, it was found that there is a correlation as shown in FIG. 7 between the annihilation time and the total neutron count regardless of the type of waste content and its density.

In this calculation simulation, net 72g U-235 is used as fissile material,
The total neutron count is proportional to the amount of fissile material. Therefore, when FIG. 7 is given per unit mass of the fissile material, the result of FIG. 8 is obtained, and the total neutron count when the fissile material is contained in the unit mass (1 g) in the radioactive solid waste can be derived from the extinction time. Based on the relationship between the annihilation time in Fig. 8 and the total count of neutrons in a unit mass (1g) of fissile material, the amount of fissile material in unknown radioactive solid waste is obtained using the procedure illustrated in Fig. 5. It can be seen that

That is, the measuring device 1 first measures the extinction time (Tx) for an unknown radioactive solid waste. Then, the measuring device 1 performs a calibration test to calculate the total neutron count (Nu) in the unit mass (1 g) of the fissile material corresponding to the annihilation time (Tx) from the actually measured annihilation time (Tx). Obtained from the obtained correlation (database, correlation journal, etc.). Next, the measuring device 1 may divide the actually measured total neutron count (Nt) by the total neutron count (Nu) per unit mass (1 g) corresponding to the measured annihilation time. Since the total count of neutrons is proportional to the fissile material, the measuring device 1 can derive or quantify the amount of fissile material contained in the measured radioactive solid waste according to the procedure illustrated in FIG. As described above, according to the measuring apparatus 1, it is possible to measure the amount of fissile material by irradiating an unknown radioactive solid waste with fast neutrons without depending on the type of the base material.

  As shown in FIGS. 1 and 2, the structure 14 serving as the outer wall of the measuring apparatus 1 includes a box-shaped fast neutron reflector 14A and a thermal neutron absorber 14B that covers the inner wall of the fast neutron reflector 14A. Therefore, thermal neutrons in the inner space of the structure 14 can be reduced. As a result, the position dependency of the measurement sensitivity of the amount of fissile material near the surface of the container 6 and the vicinity of the central axis is suppressed by using the neutron detector 12 proposed in Japanese Patent Application Laid-Open No. 2009-281878. In addition, it is further suppressed.

<Computer-readable recording medium>
A program for causing a computer or other machine or device (hereinafter, a computer or the like) to realize any of the above functions can be recorded on a recording medium that can be read by the computer or the like. The function can be provided by causing a computer or the like to read and execute the program of the recording medium.

  Here, a computer-readable recording medium is a recording medium that stores information such as data and programs by electrical, magnetic, optical, mechanical, or chemical action and can be read from a computer or the like. Say. Examples of such a recording medium that can be removed from a computer or the like include a flexible disk, a magneto-optical disk, a CD-ROM, a CD-R / W, a DVD, a Blu-ray disk, a DAT, an 8 mm tape, a flash memory, and the like. There are cards. In addition, as a recording medium fixed to a computer or the like, there are a hard disk, a ROM (read only memory), and the like.

DESCRIPTION OF SYMBOLS 1 Assumption apparatus 2 Neutron generation controller 3 Neutron pulse signal amplifier 4 Neutron pulse signal counter 5 Measurement control PC system 6 Container 11 Neutron generation source 12 Neutron detector 13 Table 14 Structure 14A Fast neutron reflector 14B Thermal neutron absorber

  The measurement method of the present invention is used in the measurement technology of fissile material by measuring clearance of nuclear fuel, or clearance verification at the time of dismantling waste disposal in a nuclear facility. The technology of the present invention also relates to a measurement algorithm for deriving the amount of fissile material in a fissile material measuring device using active neutron nondestructive measurement technology.

Claims (10)

Irradiating radioactive waste with a neutron from a fast neutron source for a predetermined time;
Measuring the number of neutrons with a detector disposed outside the radioactive waste;
The amount of fissionable material in the radioactive waste is determined based on the generation characteristics of fission neutrons generated when the neutrons incident on the radioactive waste from the neutron source cause a nuclear reaction with the fissionable material in the radioactive waste. A method for measuring the amount of fissile material. In the calculating step, out of the number of fission neutrons measured after irradiating neutrons from the neutron generation source, neutrons incident on the radioactive waste are generated by causing a nuclear reaction with the fissionable material in the radioactive waste. Calculating the number of fission neutrons generated;
Calculating the annihilation time of the fission neutron from the time variation of the number of occurrences of the fission neutron,
Calculating a total count of the number of fission neutrons generated;
The method for measuring the amount of fissile material according to claim 1, further comprising the step of deriving the amount of fissile material in the radioactive waste from the extinction time and the total count.   The step of deriving the amount of fissile material according to claim 2 is based on the calibrated correlation between the annihilation time measured in advance in the measuring apparatus used and the total count number per unit mass of the fissile material, A step of determining a total calibration count corresponding to an extinction time obtained from the radioactive waste to be measured; and a step of dividing the total count obtained from the radioactive waste to be measured by the total calibration count. Item 3. The method for measuring the amount of fissile material according to Item 2. A fast neutron source for irradiating radioactive waste with fast neutrons;
A neutron detector;
The amount of fissile material in the radioactive waste is determined based on the generation characteristics of neutrons generated by the neutron incident on the radioactive waste from the fast neutron generation source causing a nuclear reaction with the fissile material in the radioactive waste. A measuring device comprising: a calculating unit for calculating. Of the neutrons measured after irradiating neutrons from the fast neutron generation source, the arithmetic unit generates a nuclear fission in which a neutron incident on the radioactive waste causes a nuclear reaction with a fissile material in the radioactive waste Means for calculating the number of neutrons generated,
Means for calculating the annihilation time of the fission neutron from the time variation of the number of occurrences of the fission neutron,
Means for calculating the total number of fission neutrons generated;
The measurement apparatus according to claim 4, further comprising: means for deriving an amount of fissile material in the radioactive waste from the extinction time and the total count. Means for storing a correlation between the extinction time and the total number of calibrations per unit mass of the fissile material;
The means for deriving includes
Based on the correlation, a means for obtaining a total calibration count corresponding to the extinction time obtained from the radioactive waste to be measured;
The measuring apparatus according to claim 5, further comprising a unit that divides a total count obtained from the radioactive waste to be measured by the total calibration count.   The structure further comprising: a fast neutron reflector surrounding the radioactive waste, the fast neutron generation source, and the neutron detector; and a thermal neutron absorber lined on an inner wall surface of the fast neutron reflector. The measuring apparatus according to any one of 4 to 6. The fast neutron reflector includes at least one of iron, iron alloy, lead, zirconium alloy, and concrete,
The measurement apparatus according to claim 7, wherein the thermal neutron absorber includes at least one of cadmium and boron carbide. Rotating means for rotating the container containing the radioactive waste around the central axis of the container,
The measurement apparatus according to claim 4, wherein the container is disposed and rotated between the fast neutron generation source and the neutron detector. The measurement apparatus according to claim 4, further comprising a neutron moderator that covers an outer surface of the radioactive waste or surrounds the radioactive waste at a position spaced from the outer surface of the radioactive waste. .

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