JP6168582B2 - Nuclear fuel material criticality monitoring method - Google Patents

Description

  The present invention relates to a nuclear fuel material criticality monitoring method, and more particularly to a nuclear fuel material criticality monitoring method suitable for application to a nuclear power plant.

  Spent fuel assemblies used in nuclear reactors, that is, irradiation fuel assemblies loaded subcritical system for ensuring critical safety in the process of storing irradiated fuel assemblies in fuel storage racks and fuel transport storage containers An example of an effective multiplication factor measuring method is described in JP-A-5-288888. In this effective multiplication factor measurement method, the subcriticality of the subcritical system is evaluated. In other words, in the fixed source mode calculation of the neutron transport / diffusion calculation given the neutron generation rate, the geometry and group constant of the irradiated fuel assembly loaded subcritical system, a calculation with the fission cross section of the fuel region as zero was added. Using two calculated neutron flux values, the effective multiplication factor corresponding to the place where the neutron flux count rate is actually measured and evaluated at the measurement position of the neutron detector is measured and evaluated.

  The subcriticality measuring method described in JP-A-5-288888 is for a known nuclear fuel material. On the other hand, the measurement of the subcriticality of nuclear fuel material contained in the processing solution existing in the dissolved layer and piping of the nuclear fuel reprocessing facility is disclosed in Japanese Patent Laid-Open Nos. 58-45599, 62-277595, and This is described in JP-A-62-2944998.

  Japanese Patent Laid-Open No. 58-45599 discloses that a flow path in which a neutron multiplication factor is reduced is provided in parallel or in series in a pipe through which a processing liquid of a nuclear fuel reprocessing facility flows, and a neutron count rate in the flow path is used. Describes a subcriticality monitoring method that eliminates the effect of changes in the concentration of spontaneous neutron emitting nuclides in the processing liquid flowing in the pipe from the neutron counting rate of the pipe.

  In the method for monitoring the subcriticality of the nuclear fuel material existing region described in JP-A-62-277595, a neutron effective multiplication evaluation unit and a neutron flux measuring unit are formed in a pipe through which the processing liquid flows, and the neutron flux measuring unit is Surrounding the measurement neutron reflector, placing a thermal neutron absorber between the neutron flux measurement unit and the measurement neutron reflector, placing the first neutron detector outside the measurement neutron reflector, A neutron flux measuring unit structure in which a second neutron detector is disposed outside the measuring neutron reflector at a portion where a thermal neutron absorber is not disposed between the neutron flux measuring unit and the measuring neutron reflector is used. Then, the ratio of the neutron flux measured by the first neutron detector and the neutron flux measured by the second neutron detector is obtained, and the neutron flux is calculated based on the effective multiplication factor of the neutron flux measurement unit obtained by using this ratio. Monitor the subcriticality in the effective multiplication factor evaluation section.

  In the subcriticality measuring method described in Japanese Patent Application Laid-Open No. 62-294498, a dissolution tank of a nuclear fuel reprocessing facility in which a processing liquid containing a nuclear fuel material having a fissile material and a nuclide emitting spontaneous neutrons exists, Neutrons are measured at a plurality of parts having different shapes such as pipes, and the effective neutron multiplication factor is calculated based on the ratio of the neutron count rates measured at these parts.

JP-A-5-288888 JP 58-45599 A JP-A-62-277595 JP-A-62-2944998

  For example, it is assumed that an accident has occurred in a nuclear power plant and nuclear fuel material is mixed in circulating water from the reactor. In order to purify such circulating water, it is necessary to install a strainer section that prevents the inflow of nuclear fuel material or an adsorbent layer that removes radionuclides contained in the nuclear fuel material. In addition, there is a possibility that a large amount of nuclear fuel material accumulates in the bent part of the piping. When a large amount of nuclear fuel material is locally accumulated in the strainer, adsorbent layer, or bending portion where water containing nuclear fuel material flows, there is a possibility of recriticality of the nuclear fuel material. It is necessary to detect the amount of the substance.

  For the measurement of the subcriticality of such accumulated nuclear fuel material, the subcriticality described in JP-A-58-45599, JP-A-62-277595, or JP-A-62-2944998 is used. It is conceivable to apply a measurement method. These subcriticality measurement methods measure the neutron count rate and measure the subcriticality of the nuclear fuel material accumulated based on the measured neutron count rate. However, based on the neutron count rate, the criticality of nuclear fuel material accumulated at a certain location in the nuclear power plant cannot be monitored with high accuracy.

  An object of the present invention is to provide a criticality monitoring method for nuclear fuel material that can more accurately monitor the subcritical state of nuclear fuel material accumulated in a system in which a liquid containing nuclear fuel material flows.

  The feature of the present invention that achieves the above-described object is obtained by obtaining a first amount of nuclear fuel material accumulated in a nuclear power plant system, detecting neutrons emitted from the nuclear fuel material, and detecting the neutrons. Using the neutron count rate, the data processor uses the neutron inverse multiplication method to determine the reciprocal of the effective neutron multiplication factor, and the data processor determines the critical quantity based on the first quantity determined and the inverse of the effective neutron multiplication factor. The second quantity of nuclear fuel material accumulated at the point is to be calculated.

  Since the second amount of nuclear fuel material accumulated at the critical point is calculated based on the obtained first amount of nuclear fuel material and the reciprocal of the effective neutron multiplication factor, the second amount of nuclear fuel material can be accurately calculated. . For this reason, the subcritical state of the nuclear fuel material accumulated in the system through which the liquid containing the nuclear fuel material flows can be monitored with higher accuracy.

  ADVANTAGE OF THE INVENTION According to this invention, the subcritical state of the nuclear fuel material accumulate | stored in the system | strain in which the liquid containing a nuclear fuel material flows can be monitored more accurately.

It is a block diagram of the criticality monitoring apparatus used for the criticality monitoring method of the nuclear fuel material of Example 1 which is one preferable Example of this invention. It is explanatory drawing which shows an example of the gamma ray spectrum distribution of a nuclear fuel substance. It is a characteristic view which shows the relationship of the neutron count rate with respect to the weight of a nuclear fuel material. It is explanatory drawing which shows the concept of the neutron reverse multiplication method applied to Example 1. FIG. It is a block diagram of the criticality monitoring apparatus used for the criticality monitoring method of the nuclear fuel material of Example 2 which is another Example of this invention. It is a block diagram of the criticality monitoring apparatus used for the criticality monitoring method of the nuclear fuel material of Example 3 which is another Example of this invention.

  Examples of the present invention will be described below.

  A criticality monitoring method for nuclear fuel material according to embodiment 1, which is a preferred embodiment of the present invention, will be described with reference to FIG.

The criticality monitoring device 1 used for the nuclear fuel material criticality monitoring method of the present embodiment includes a He-3 proportional counter (neutron detection device) 2, a LaBr ₃ (Ce) detector (γ-ray detection device) 3, and a data processing device. 4 and the display device 5. The He-3 proportional counter 2 and the LaBr ₃ (Ce) detector 3 are connected to the data processing device 4. The display device 5 is connected to the data processing device 4. As a γ-ray detection device, a semiconductor radiation detector such as a CdTe semiconductor detector, a Ge (Li) semiconductor detector, or a scintillator such as a NaI (Tl) detector may be used instead of the LaBr ₃ (Ce) detector. good.

  In the nuclear power plant, a strainer 11 is installed in the piping 10 assuming that an accident occurs in which nuclear fuel material flows into the cooling water 12 flowing in the piping 10 connected to the nuclear reactor. When the above accident occurs, the nuclear fuel material 13 contained in the cooling water 12 flowing through the pipe 10 is removed by the strainer 11. The removed nuclear fuel material 13 is accumulated in the strainer 11. Instead of the strainer 11, an adsorbent layer that adsorbs the radionuclide contained in the nuclear fuel material may be provided in the pipe 10.

The criticality monitoring device 1 is a device that monitors the criticality of the nuclear fuel material 13 removed by the strainer 11 and accumulated in the strainer 11. The He-3 proportional counter 2 and the LaBr ₃ (Ce) detector 3 are connected to the strainer 11. It is arranged outside the pipe 10 at the installation position. When an accident occurs in which nuclear fuel material flows into the cooling water 12 flowing in the pipe 10, the power switch connected to the criticality monitoring device 1 is turned on, the He-3 proportional counter 2, the LaBr ₃ (Ce) detector 3. A current is supplied to the data processing device 4 and the display device 5. For this reason, the He-3 proportional counter 2, the LaBr ₃ (Ce) detector 3, the data processing device 4, and the display device 5 are activated.

  If the weight of the nuclear fuel material accumulated in the strainer 11 increases, the nuclear fuel material in the strainer 11 may become locally recritical. For this reason, it is necessary to predict critical proximity within the strainer 11, and the criticality of the nuclear fuel material accumulated in the strainer 11 is monitored by the criticality monitoring device 1. The measurement of subcriticality using the criticality monitoring apparatus 1 will be described in detail below.

The γ rays emitted from the nuclear fuel material 13 accumulated in the strainer 11 are detected by the LaBr ₃ (Ce) detector 3. The LaBr ₃ (Ce) detector 3 outputs a γ-ray detection signal by detecting γ-rays. This γ-ray detection signal is input to the data processing device 4. The data processing device 4 obtains a count rate (γ-ray count rate) of the input γ-ray detection signal, and obtains a γ-ray spectrum distribution using the obtained γ-ray count rate. An example of this γ-ray spectrum distribution is shown in FIG. The γ-ray spectrum distribution includes a γ-ray count rate for each peak corresponding to the component materials of the nuclear fuel material accumulated in the strainer 11, for example, Cs-137, Cs-134, and Eu-154. This gamma ray spectral distribution further includes a pulser signal 20.

  Based on the γ-ray spectrum distribution, the data processing device 4 calculates the weight of the accumulated nuclear fuel material, which is an index used for predicting critical proximity together with a neutron count rate described later. The calculation of the weight of the nuclear fuel material is obtained based on the γ-ray count rate (for example, Cs-137) of a certain peak in the γ-ray spectrum distribution. When determining the weight of the nuclear fuel material 13 accumulated in the strainer 11 based on the γ-ray count rate of the Cs-137 peak, the weight determined based on the γ-ray count rate of the Cs-137 peak is What is necessary is just to divide by the ratio of Cs-137 contained in the nuclear fuel material 13.

  Neutrons emitted from the nuclear fuel material accumulated in the strainer 11 are detected by the He-3 proportional counter 2. This neutron is mainly generated by spontaneous fission of curium (Cm) contained in nuclear fuel material. The He-3 proportional counter 2 outputs a neutron detection signal by detecting neutrons. The data processing device 4 obtains a neutron count rate based on the input neutron detection signal. This neutron count rate is small at the initial stage when nuclear fuel material is accumulated in the strainer 11 and increases in proportion to the weight of the nuclear fuel material accumulated in the strainer 11 (see FIG. 3). When the weight of the nuclear fuel material accumulated in the strainer 11 increases over time to a certain amount (the weight of the nuclear fuel material at the position of the arrow 22 shown in FIG. 3), the neutron count rate is the effect of multiplication. As shown in FIG. 3, it increases rapidly.

  Immediately after the above accident has occurred and the He-3 proportional counter 2 has been started, that is, in the initial stage described above, the state in which the neutron count rate is small continues as described above. Based on the neutron detection signal output from the He-3 proportional counter 2 when the nuclear fuel material accumulated in the strainer 11 increases to the weight of the nuclear fuel material indicated by the arrow 22 and enters a critical proximity state. The obtained neutron counting rate is increased. For this reason, the data processing device 4 uses the statistics for averaging the neutron count rate when the neutron count rate is lower than the neutron count rate (hereinafter referred to as the set neutron count rate) corresponding to the weight of the nuclear fuel material indicated by the arrow 22. When the time is long and the neutron count rate is higher than the set neutron count rate, the statistical time is shortened. Thereby, the statistical error of the average value of the neutron count rate calculated | required with the data processor 4 can be made comparable, and the quick response at a high count rate can be ensured. The set neutron count rate is the value of the neutron count rate at the time when the degree of increase in the neutron count rate changes as the weight of the nuclear fuel material increases.

  The data processing device 4 obtains the critical point of the nuclear fuel material 13 accumulated in the strainer 11 by the neutron reverse multiplication method using the obtained neutron count rate.

  A general neutron inverse multiplication method uses an inverse number (1 / M) of an effective neutron multiplication factor calculated by using a neutron count rate obtained based on a neutron detection signal output from a neutron measurement device, for example, Using the weight of nuclear fuel material loaded in the core as an index, as shown in FIG. 4, a curve (hereinafter referred to as a reverse multiplication curve) is obtained by plotting each measurement step (step 1 and step 2 etc.). This is a method of estimating the weight (predicted critical amount) of nuclear fuel material necessary to become critical (1 / M = 0) by extrapolating the inverse multiplication curve (see FIG. 4). By using the neutron reverse multiplication method, the margin to the critical point (1 / M = 0) can be calculated and predicted quantitatively before reaching the supercriticality, thereby avoiding the risk of unexpected supercriticality. can do. Conventionally, clear measured value data such as the weight and concentration of nuclear fuel material, the amount of moderator and reflector, the position of the control rod, and the concentration of boric acid have been used as an index of the reciprocal of the effective neutron multiplication factor.

In the present embodiment, the information indicating the weight of the nuclear fuel material 13 accumulated in the strainer 11, for example, the γ-ray spectrum distribution obtained using the γ-ray detection signal output from the LaBr ₃ (Ce) detector 3 is used. The γ-ray count rate of a certain peak included (for example, the γ-ray count rate of the Cs-137 peak) is used. In this embodiment, since the curium, which is a spontaneous fissile material contained in the accumulated nuclear fuel material 13, is used as a neutron source without using an external neutron source, the intensity of the neutron source varies.

  The neutron reverse multiplication method used in the present embodiment will be described. When the nuclear fuel material 13 contained in the cooling water 12 flowing in the pipe 10 is captured by the strainer 11, the weight of the nuclear fuel material 13 accumulated in the strainer 11 increases, and accordingly, the nuclear fuel material in the strainer 11. The weight of curium (neutron source) contained in 13 also increases. For this reason, since the intensity | strength of the neutron source (curium) contained in the nuclear fuel material 13 in the strainer 11 changes, the neutron reverse multiplication method used in a present Example is the general neutron reverse multiplication method mentioned above. The principle is the same, but the calculation formula is different. The calculation formula of the neutron reverse multiplication method used in this embodiment will be described below.

Between the neutron counting rate C and the effective neutron multiplication factor k _eff , there is a relationship represented by the formula (1).

Here, S is the intensity of the neutron source contained in the accumulated nuclear fuel material at a certain measurement time t, and ε is the neutron detection efficiency.

Assuming that the neutron count rate C ₀ is obtained at the starting time t ₀ (for example, the neutron measurement start time), and the neutron multiplication factor at this time is k ₀ , the neutron count rate C ₀ is given by equation (2). expressed.

Here, S ₀ is the intensity of the neutron source contained in the accumulated nuclear fuel material at the starting point t ₀ (for example, the neutron measurement start point). The neutron detection efficiency ε is assumed to be constant on the assumption that the measurement conditions of the neutron detector are not changed. By transforming equation (2), equation (3) is obtained.

1 / M is expressed as equation (4) by modifying equation (3).

The calculation of the reciprocal 1 / M of the neutron effective multiplication factor using the neutron count rate obtained in the data processor 4 based on the neutron detection signal output from the He-3 proportional counter 2 will be described. The neutron count rate C in the equation (4) is determined by the data processing device 4 based on the neutron detection signal output from the He-3 proportional counter 2 at a certain measurement time t after a lapse of time from the starting time t _0. The obtained neutron count rate is substituted, and the data processor 4 is based on the neutron detection signal output from the He-3 proportional counter 2 at the starting time t ₀ to the neutron count rate C ₀ in the equation (4). Substituting the neutron counting rate obtained by In the conventional method, in order to use a fixed radiation source, a fixed value is substituted into the neutron source intensity S at a certain measurement time point t after the time point t ₀ as the starting point shown in the equation (4). It was. However, in this embodiment, the neutron source intensity S is one of the two methods described below in order to use the spontaneous fissionable material contained in the nuclear fuel material 13 whose concentration is not yet determined in the strainer 11 as a neutron source. Is calculated by

The first method for determining the neutron source strength S is at time t ₀ as the starting point, to the intensity S ₀ of the neutron source contained in the stored nuclear fuel material, some measurement time has elapsed from the time t ₀ as the starting point This is a method of approximating the ratio (S / S ₀ ) of the intensity S of the neutron source contained in the accumulated nuclear fuel material at time t to a constant (S / S ₀ = 1). That is, the first method assumes that the intensity of the accumulated neutron source in the nuclear fuel material 13 is constant at the measurement start time (starting time t ₀ ) and at a certain measurement time t after this start time. This is a method of eliminating the unknown neutron source intensity S ₀ and neutron source intensity S in equation (4).

  As shown in Equation (5), the neutron source intensity S is proportional to the weight of the accumulated nuclear fuel material 13, but increases rapidly in the vicinity of the criticality (position of the arrow 22 in FIG. 3). Changes in the neutron source intensity S can be ignored compared to changes in the neutron count rate C.

The second method for obtaining the neutron source intensity S is a method that approximates that it is proportional to the peak value of γ rays of a certain radionuclide (for example, Cs-137) contained in the nuclear fuel material 13 in which the neutron source intensity S is accumulated. is there. The accumulated nuclear fuel material 13 in the γ-ray spectrum distribution obtained as the nuclear fuel material 13 contained in the cooling water 12 is accumulated in the strainer 11 provided in the pipe 10 through which the cooling water 12 flows. The peak of the counting rate of γ rays emitted from the radionuclides contained in is increasing. Similarly, the neutron source intensity S increases. Therefore, it is assumed that there is a relationship of S = αγ _max where γ _{max is} the peak value of the γ-ray count rate at a certain measurement time t and α is a proportionality coefficient. In this case, S / S ₀ in Equation (4) is obtained as S / S ₀ = αγ _max / αγ _0max = γ _max / γ _0max . That is, S / S ₀ is calculated from the peak value of γ rays.

The neutron intensity S cannot be simply estimated based on the neutron count rate C obtained based on the neutron detection signal that is the output of the neutron measurement device. The reason for this will be described below. The neutron intensity S increases in proportion to the accumulation amount of nuclear fuel material contained in the cooling water 12, but the neutron effective multiplication factor k _eff also increases due to the accumulation. When the effective neutron multiplication factor k _eff increases and approaches 1, the 1 / (1-k _eff ) term of equation (5) becomes 1000 times, 10000 times, and so on. Therefore, the neutron count rate C is the effective neutron multiplication factor. It is greatly influenced by the value of k _eff . Therefore, the neutron counting rate C and the neutron intensity S cannot be simply associated.

The reciprocal 1 / M of each neutron effective multiplication factor at a plurality of measurement points in time t is calculated using Equation (4). For example, it is assumed that the reciprocals 1 / M at measurement time points t ₁ , t _2, and t ₃ with different elapsed times from the time point t ₀ that is the base point that is the measurement start time point of neutrons. Reciprocal 1 / _{M 1} in the measurement time _{t 1,} the neutron multiplication factor of the value of the neutron count rate _{C 0} at time _{t 0} to _{C 0} of the formula (4), at time _{t 0} to _{k 0} of the formula (4) The value of k _{0 and} the value of the neutron count rate C ₁ at the measurement time t ₁ are substituted into C in the equation (4), respectively. The neutron multiplication factor k ₀ is a neutron multiplication factor at the measurement start time (time t ₀ ), and it is assumed that k ₀ = 0. In this embodiment, since S / S _{0 is obtained} by the second method, for example, the peak value γ _0max of the γ-ray count rate of Cs-137 at the time point t ₀ and the γ-ray of Cs-137 at the measurement time point t _1. Γ _1max / γ _0max is calculated using the peak value γ _1max of the count rate. This calculation is performed by the data processing device 4. The value of γ _1max / γ _0max is substituted into S / S ₀ in equation (4). The data processing device 4 calculates the reciprocal 1 / M ₁ of the effective neutron multiplication factor at the measurement time t ₁ by the equation (4) in which these values are substituted.

The reciprocal 1 / M ₂ of the effective neutron multiplication factor at the measurement time t ₂ is calculated by substituting the neutron count rate C ₀ and the neutron multiplication factor k ₀ into equation (4) in the same manner as calculating the reciprocal 1 / M _1. C to the value of the neutron count rate _{C 2} at the measurement time _{t 2} by substituting each), substitutes the value of gamma _2max / gamma _0max the S / _{S 0.} γ _2max is the peak value of the γ-ray count rate of Cs-137 at the measurement time t ₂ . Reciprocal 1 / _{M 3} of effective neutron multiplication factor at the measurement time _{t 3} likewise, substituting Equation (4) to the value of the value of the neutron count rate _{C 3} at the measurement time _{t 3} and gamma _3max / gamma _0max etc. Is calculated by gamma _3max is the peak value of the gamma ray counting rate of Cs-137 in the measurement time _{t 2.}

The reciprocal 1 / M ₁ is the value of the reciprocal of the effective neutron multiplication factor in step 1 (see FIG. 4). Similarly, the reciprocal 1 / M ₂ is the value of the reciprocal of the effective neutron multiplication factor in step 2 and 1 / M ₃ is a reciprocal value of the effective neutron multiplication factor in step 3. Measuring time t ₃ may calculate the value of the inverse 1 / M of the respective effective neutron multiplication factor in the subsequent measurement time point t ₄ and t _5, and the like.

Further, the data processing device 4 determines the weights W ₀ , W ₁ and W ₂ of the nuclear fuel material 13 accumulated in the strainer 11 at the measurement start time t ₀ and the measurement times t ₁ and t ₂ , respectively. As described above, the calculation is based on the γ-ray count rate of the Cs-137 peak at each time point.

For example, the data processing device 4 is configured such that each of the reciprocals 1 / M ₀ , 1 / M ₁ and 1 / M ₂ of the effective neutron multiplication factor, and the weights W ₀ , W ₁ and W ₂ of the nuclear fuel material 13 are obtained. Using the values, an equation of a reverse multiplication curve (curve indicated by a solid line in FIG. 4) showing the relationship between the weight of the nuclear fuel material 13 and the reciprocal of the neutron effective multiplication factor is obtained by the least square method. Using the obtained inverse multiplication curve equation, the data processing device 4 calculates the weight of the nuclear fuel material 13 accumulated in the strainer 11 when the reciprocal 1 / M = 0 of the effective neutron multiplication factor. . Weight of the nuclear fuel material 13 when it comes to the reciprocal 1 / M = 0 the effective neutron multiplication factor, the weight W _C of the nuclear fuel material 13 at the critical point.

The weight W _C of the nuclear fuel material 13 obtained by the data processing device 4 and the information on the weights W ₁ and W ₂ of the nuclear fuel material accumulated in the strainer 11 at the measurement times t1 and t2 before the critical point are respectively obtained. The data is output from the data processing device 4 to the display device 5 and displayed on the display device 5. The operator can confirm the subcritical state of the nuclear fuel material 13 accumulated in the strainer 11 by looking at the displayed weight of each nuclear fuel material. The data processing device 4 is configured so that the weight of the nuclear fuel material 13 accumulated in the strainer 11 is set with a margin of the nuclear fuel material weight W _{C at} the critical point (the weight W of the nuclear fuel material). when increased to smaller) than _C, it generates alarm information, and displays the alarm information to the display device 5. The alarm information may be audio information and output to a speaker.

  Further, alarm information output from the data processing device 4 is input to a control device (not shown), which opens a valve provided in a boron solution injection pipe connected to a tank storing boron solution. Then, an infusion pump provided in the boron solution infusion tube is driven. By these operations, the boron solution in the tank is injected into the pipe 10, and the boron 10 contained in the boron solution absorbs neutrons released from the nuclear fuel material 13 accumulated in the strainer 11. For this reason, it can avoid that the nuclear fuel material 13 accumulate | stored in the strainer 11 will be in a critical state. Further, based on the output alarm information, the control device stops the operation of a pump (not shown) provided in the pipe 10 and stops the supply of cooling water to the strainer 11. Further, when alarm information is generated, a valve (not shown) provided in the pipe 10 may be fully closed.

In the present embodiment, even if an accident that the nuclear fuel material 13 flows out into the cooling water 12 occurs, the location of the nuclear fuel material 13 contained in the cooling water 12 in the nuclear power plant (for example, provided in the pipe 10). The LaBr ₃ (Ce) detector 3 detects γ-rays released from the nuclear fuel material 13 accumulated in the strainer 11), and based on the γ-ray detection signal output from the LaBr ₃ (Ce) detector 3. Information on the γ-ray spectrum distribution is created using the obtained γ-ray count rate, and the peak γ-ray count rate of a certain radionuclide (for example, Cs-137) included in the γ-ray spectrum distribution information is used. The weight of the nuclear fuel material 13 accumulated in the strainer 11 is obtained. Furthermore, in this embodiment, neutrons generated by spontaneous fission of spontaneous fissionable material released from nuclear fuel material 13 accumulated in the strainer 11 are detected by the He-3 proportional counter 2, and the He-3 proportional count is detected. A neutron count rate is determined based on the neutron detection signal output from the tube 2.

Further, as described above, the value of the neutron count rate obtained corresponding to each measurement time point is substituted into the equation (4), that is, the neutron effective multiplication at each measurement time point by the neutron inverse multiplication method. The reciprocal 1 / M of the magnification is calculated. The reciprocal 1 / M of the effective neutron multiplication factor 1 / M and the reciprocal 1 / M of the effective neutron multiplication factor 0 based on the calculated nuclear fuel material weight at each measurement time (critical point) ) weight W _C of the nuclear fuel material accumulated in is required.

Thus, according to the present embodiment, based on the weight of the nuclear fuel material 13 accumulated in the strainer 11 and the reciprocal 1 / M of the neutron effective multiplication factor obtained using the neutron counting rate, weight W _C of the nuclear fuel material accumulated can be determined more accurately. For this reason, even if an accident that the nuclear fuel material 13 flows into the cooling water 12 occurs, this embodiment accumulates the nuclear fuel material 13 contained in the cooling water 12 in the nuclear power plant system. Thus, the subcritical state of the nuclear fuel material 13 can be monitored with higher accuracy.

In this embodiment, the LaBr ₃ (Ce) detector 3 detects γ rays emitted from the nuclear fuel material 13 accumulated in the strainer 11 provided in the pipe 10, and outputs from the LaBr ₃ (Ce) detector 3. Γ-ray spectrum distribution information is generated using the γ-ray count rate obtained based on the detected γ-ray detection signal, and a certain radionuclide (for example, Cs-137) included in the γ-ray spectrum distribution information is created. Since the weight of the nuclear fuel material 13 accumulated in the strainer 11 is obtained by using the peak γ-ray count rate, the accumulated amount changes with the passage of time, and the accumulated amount of the nuclear fuel material 13 in the strainer 11 is unknown. Can be grasped with higher accuracy.

  When the conventional critical neutron inverse multiplication method is applied and the predicted critical amount of nuclear fuel material is estimated by extrapolation of the inverse multiplication curve, only the measured value of the neutron count rate is used. Even if there was a multiplication of the neutron, there was no choice but to predict that it would become critical based on a rapid rise in the neutron count rate. For this reason, in the conventional example, the criticality of the nuclear fuel material accumulated in a certain part of the nuclear power plant could not be accurately monitored.

In this embodiment, the neutron count rate measured by detecting the neutrons emitted from the accumulated nuclear fuel material, and the weight of the accumulated nuclear fuel material (in this embodiment, LaBr ₃ (Ce) detection, which is a γ-ray detector). The criticality of the accumulated nuclear fuel material can be accurately predicted using the weight of the nuclear fuel material calculated based on the γ-ray detection signal output from the vessel 3, so that the subcriticality of the accumulated nuclear fuel material can be predicted. The state can be monitored with higher accuracy.

  In Examples 2 and 3 to be described later, the weight of the mixture contained in the nuclear fuel material is determined together with the weight of the nuclear fuel material. In the present embodiment in which the weight of the nuclear fuel material is obtained based on the γ-rays emitted from the accumulated nuclear fuel material, the weight of the accumulated nuclear fuel material can be obtained with higher accuracy than in Examples 2 and 3 described later. . For this reason, the present embodiment can monitor the subcritical state of the accumulated nuclear fuel material more accurately than the second and third embodiments.

  A criticality monitoring method for nuclear fuel material according to embodiment 2, which is another embodiment of the present invention, will be described with reference to FIG.

The criticality monitoring device 1A used in the nuclear fuel material criticality monitoring method of the present embodiment replaces the LaBr ₃ (Ce) detector 3 with a load cell (load measuring device) 6 in the criticality monitoring device 1 used in the first embodiment. It has a configuration. The load cell 6 is connected to the data processing device 4. Other configurations of the criticality monitoring device 1A are the same as those of the criticality monitoring device 1.

  The load cell 6 is installed in a canister 14 as a container, and a suction hose 15 for transferring nuclear fuel material is connected to the canister 14.

  In a nuclear power plant, an accident that nuclear fuel material flows out from the fuel rods loaded in the reactor core inside the reactor pressure vessel into the cooling water should occur, and the nuclear fuel material 16 is present at the location where the cooling water flows in the nuclear power plant. Assume the case of accumulation. The nuclear fuel material 16 accumulated in the place is transferred into the canister 14 together with the cooling water through the suction hose 15. A strainer (not shown) is installed in the canister 14, and cooling water is separated by the strainer. As a result, the nuclear fuel material 16 accumulates on the load cell 6. The load cell 6 measures the weight of the accumulated nuclear fuel material 13. Information on the weight of the nuclear fuel material 13 measured by the load cell 6 is input to the data processing device 4.

When the weight of the nuclear fuel material 16 is measured by the load cell 6, the He-3 proportional counter 2 detects neutrons emitted from the spontaneous fissile material (for example, curium) contained in the nuclear fuel material 13 in the canister 14. And outputs a neutron detection signal. This neutron detection signal is input to the data processing device 4. The data processing device 4 obtains the neutron count rate based on the neutron detection signal, and substitutes the neutron count rate and the like into the equation (4) and uses the same as in Example 1, using the reciprocal 1 / M of the neutron effective multiplication factor. Ask for. Further, the data processing device 4 obtains an inverse multiplication curve equation based on the weight of each nuclear fuel material 13 and the inverse 1 / M of each neutron effective multiplication factor, and using this inverse multiplication curve equation, the neutron effective multiplication factor is obtained. to calculate the weight W _C of the nuclear fuel material at the critical point of the reciprocal 1 / M magnification becomes zero. Weight W _C and weight information of the nuclear fuel material at each measurement time point of the calculated nuclear fuel material is displayed on the display device 5.

  In the present embodiment, each effect produced in the first embodiment can be obtained.

  A criticality monitoring method for nuclear fuel material according to embodiment 3, which is another embodiment of the present invention, will be described with reference to FIG.

A criticality monitoring device 1B used in the nuclear fuel material criticality monitoring method of the present embodiment has a configuration in which the LaBr ₃ (Ce) detector 3 is replaced with a differential pressure gauge 7 in the criticality monitoring device 1 used in the first embodiment. Other configurations of the criticality monitoring device 1A are the same as those of the criticality monitoring device 1.

  The differential pressure gauge 7 is connected to the data processing device 4. A pressure pipe that opens into the pipe 10 upstream from the strainer 11 provided in the pipe 10 and another pressure pipe that opens into the pipe 10 downstream from the strainer 11 provided in the pipe 10 are a pressure gauge. 7 is connected.

  In the unlikely event that an accident occurs in which the nuclear fuel material 13 flows into the cooling water 12, the cooling water 12 containing the nuclear fuel material 13 flows through the pipe 10. Since the cooling water 12 passes through the strainer 11, the nuclear fuel material 13 contained in the cooling water 12 is captured by the strainer 11 and accumulated in the strainer 11 as in the first embodiment. When the nuclear fuel material 13 is accumulated in the strainer 11, a pressure difference is generated between the upstream of the strainer 11 and the downstream of the strainer 11 corresponding to the accumulated amount of the nuclear fuel material 13. The differential pressure gauge 7 detects the difference between the pressure in the pipe 10 on the upstream side of the strainer 11 and the pressure in the pipe 10 on the downstream side of the strainer 11, and outputs information on the detected differential pressure to the data processing device 4. . The differential pressure measured by the differential pressure gauge 7 is correlated with the weight of the nuclear fuel material 13 accumulated in the strainer 11. When the weight of the accumulated nuclear fuel material 13 increases, the differential pressure measured by the differential pressure gauge 7 also increases. The data processing device 4 calculates the weight of the accumulated nuclear fuel material based on the input differential pressure measurement value.

When the pressure difference between the upstream side and the downstream side of the strainer 11 is measured by the differential pressure gauge 7, the He-3 proportional counter 2 is a spontaneous fissile material (for example, curium contained in the nuclear fuel material 13 accumulated in the strainer 11. ) Is detected and a neutron detection signal is output. This neutron detection signal is input to the data processing device 4. The data processing device 4 obtains the neutron count rate based on the neutron detection signal, and substitutes the neutron count rate and the like into the equation (4) and uses the same as in Example 1, using the reciprocal 1 / M of the neutron effective multiplication factor. Ask for. Further, the data processing device 4 obtains the formula of the reverse multiplication curve based on the calculated weight of each nuclear fuel material 13 and the reciprocal 1 / M of each neutron effective multiplication factor, and the neutron using the formula of the reverse multiplication curve. reciprocal 1 / M of the effective multiplication factor to calculate the weight W _C of the nuclear fuel material at the critical point becomes zero. Weight W _C and weight information of the nuclear fuel material at each measurement time point of the calculated nuclear fuel material is displayed on the display device 5.

In the present embodiment, each effect produced in the first embodiment can be obtained. In the present embodiment, since the weight of the nuclear fuel material is obtained based on the differential pressure measured by the differential pressure gauge 7, the γ-ray detector (for example, LaBr ₃ (Ce) detector) in the first embodiment, and the second embodiment. The load cell 6 and the canister 14 are not required, and the weight of the nuclear fuel material accumulated with a simple configuration can be obtained.

1, 1A, 1B ... critical monitoring device, 2 ... the He-3 proportional counter (neutron detector), 3 ... LaBr ₃ (Ce) detector (gamma ray detector), 4 ... data processing unit, 6 ... load cell, 7 ... differential pressure gauge, 10 ... piping, 11 ... strainer, 12 ... cooling water, 13 ... nuclear fuel material, 14 ... canister.

Claims (8)

Detects gamma rays emitted from nuclear fuel material accumulated in the nuclear power plant system,
Create γ-ray spectrum information using the γ-ray count rate obtained by detecting the γ-ray,
First information representing the weight of the accumulated nuclear fuel material is obtained based on a gamma ray count rate of a peak of Eu-154 included in the gamma ray spectrum information,
Detecting neutrons emitted from the nuclear fuel material;
Using the neutron count rate obtained by the detection of the neutron, in the data processing device, by the neutron reverse multiplication method, obtain the reciprocal of the neutron effective multiplication factor,
As the nuclear fuel material is accumulated in the system, a plurality of sets of first information representing the weight and reciprocals of the effective neutron multiplication factor corresponding to the first information representing the weight are obtained,
Wherein the data processing unit, based on the inverse of the first information及beauty the effective neutron multiplication factor representing said plurality of sets of weights determined, the relationship between the reciprocal of the first information indicating the weight the effective neutron multiplication factor Find the reverse multiplication curve that represents
The inverse multiplication curve using critical monitoring method of the nuclear fuel material, characterized in Rukoto seek second information representative of the weight of the nuclear fuel material is stored in a critical point. The first information representing the weight of the accumulated nuclear fuel material is the first weight of the accumulated nuclear fuel material;
2. The method for monitoring criticality of nuclear fuel material according to claim 1, wherein the second information indicating the weight of the nuclear fuel material accumulated at the critical point is a second weight of the nuclear fuel material accumulated at the critical point. The first weight is obtained by determining the weight of Eu-154 based on the γ-ray count rate of the peak of Eu-154 included in the γ-ray spectrum information, and the obtained weight of Eu-154 is included in the nuclear fuel material. The method for monitoring criticality of nuclear fuel material according to claim 2 , obtained by dividing by the ratio of Eu-154. The first by weight of said nuclear fuel material obtained at a certain measurement point, the nuclear fuel the determination section determines whether less than the set weight of the said nuclear fuel material less than the double amount of material 2 or 3 The criticality monitoring method for nuclear fuel materials described in 1. When the first by weight is equal to or greater than the setting Weight critical monitoring method of the nuclear fuel material according to claim 4, the alarm information is generated. When said first by weight is equal to or greater than the setting Weight critical monitoring method of the nuclear fuel material according to claim 4 for injecting a solution containing a neutron absorbing material in said system, wherein the nuclear fuel material is stored. The nuclear fuel material according to any one of claims 1 to 6, wherein the accumulated nuclear fuel material is nuclear fuel material leaked from the fuel rods loaded in the reactor pressure vessel into the cooling water in the reactor pressure vessel. Criticality monitoring method for substances. A first statistical time for averaging the neutron count rate when the neutron count rate is less than or equal to a set neutron count rate, and averaging the neutron count rate when the neutron count rate is greater than the set neutron count rate The method for monitoring criticality of nuclear fuel material according to any one of claims 1 to 7 , wherein the critical time is longer than the second statistical time.

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