Apparatus for measuring subcriticality of spent nuclear fuel and method for managing criticality of spent nuclear fuel

Abstract

To provide a spent nuclear fuel degree-of-subcriticality measurement device and the like that achieve cost reduction, and enable an estimation of a neutron multiplication factor to be implemented as a function of a second step.SOLUTION: A spent nuclear fuel degree-of-subcriticality measurement device, which performs measurement of a first step of measuring the number of neutrons of a spent nuclear fuel to measure the degree of burn-up, and measurement of a second step with respect to the spent nuclear fuel not meeting the degree of burn-up limit value in the first step, comprises: accumulation means of accumulating the number of neutrons as time-series data; time-series data analysis means of calculating a prompt neutron attenuation constant α from the time-series data; pre-analysis means of calculating a delayed neutron ratio β and a prompt neutron lifetime L by preliminarily neutron transport analysis; and neutron multiplication factor means of calculating a neutron multiplication factor k from the prompt neutron attenuation constant α, delayed neutron ratio β and prompt neutron lifetime L. The degree of subcriticality of the second step is determined by the neutron multiplication factor k.SELECTED DRAWING: Figure 1

Description

[Technical field]

[0001] An embodiment of the present invention relates to a subcriticality measuring device for spent nuclear fuel and a method for criticality management of spent nuclear fuel. [Background technology]

[0002] In general, spent nuclear fuel from nuclear power plants is cooled for a specified period of time, then stored in a transport container and transported to an intermediate storage facility or a reprocessing facility. The spent nuclear fuel sent to this intermediate storage facility is further cooled for several decades before being permanently disposed of or sent to a reprocessing facility.

[0003] At the reprocessing facility, the spent nuclear fuel is removed from the transport container and verified to have been transported correctly by comparing the transport data from the power plant with the actual item, or by non-destructive testing of burnup, etc. This spent nuclear fuel is then stored for an appropriate period of time, after which it is reprocessed according to a set schedule.

[0004] Spent uranium fuel initially contains a large amount of fissile uranium 235. The amount of uranium 235 initially contained is called the initial enrichment, and on average a fuel assembly is enriched to about 3-4% uranium 235, but in spent uranium fuel this has decreased to about 1-2% due to depletion caused by uranium fission. However, spent uranium fuel still retains the ability to undergo nuclear fission, and criticality safety management is required in spent fuel reprocessing plants to prevent spent uranium fuel from going "critical."

[0005] The most conservative method of criticality safety management is called the "fresh fuel assumption," which assumes that the spent nuclear fuel is not burning and determines the placement of the spent nuclear fuel and the placement of neutron absorbing materials that suppress nuclear fission, but this increases costs. Therefore, a more rational method of criticality safety management that considers that the reactivity of spent nuclear fuel decreases as the fuel burns is called "burnup credit."

[0006] In order to adopt burnup credits, it is necessary to know the burnup of the fuel. Methods and devices for non-destructively evaluating the burnup of the fuel based on gamma ray or neutron measurements have been known (see, for example, Patent Documents 1 and 2).

[0007] The neutron multiplication factor k is used as an index of the susceptibility of fuel to criticality. The neutron multiplication factor k is defined as (neutron production) / (neutron annihilation). If it is assumed that the fuel assemblies are arranged infinitely, the neutron multiplication factor k is known as the infinite multiplication factor k∞. In addition, the neutron multiplication factor k in a real finite system is the effective multiplication factor k eff Hereafter, the neutron multiplication factor k is the effective multiplication factor k eff represents the following.

[0008] As the fuel is burned, the enrichment of uranium decreases, and as this infinite multiplication factor k∞ decreases, it becomes less likely to go critical. In conventional spent uranium fuel reprocessing plants, the infinite multiplication factor k∞ must be reduced to a certain value, i.e., the fuel must be burned at a certain burnup or higher, and criticality safety design is carried out based on the reduced uranium enrichment (residual enrichment) that corresponds to this burnup. This is called residual enrichment management. It is difficult to assess residual uranium enrichment non-destructively, so as mentioned above, the measurable burnup is assessed using gamma ray and neutron measurements, and the residual enrichment is calculated from the burnup.

[0009] As mentioned above, burnup is calculated from gamma rays and neutron measurements, but the generation of gamma rays and neutrons that accompany the burning of nuclear fuel depends on the operating history of the reactor. For example, in a boiling water reactor (BWR), bubbles (voids) are generated inside the reactor. There is less void generation in the lower part of the fuel. On the other hand, there is more void generation in the upper part of the fuel. It is known that in the high void fraction in the upper part, many neutrons are generated even with the same burnup.

[0010] In addition, the risk of fuel with an unexpected burnup being mixed in due to incorrect loading cannot be completely eliminated. For this reason, the burnup measurement device is designed to measure the neutrons in the first step due to its unique operating history. totalThe second step is prepared as a backup measuring device for fuel that indicates a value that does not satisfy the burnup limit value in the burnup measurement by the number rate measurement. At the Rokkasho Reprocessing Plant, in the second step, neutron multiplication factor measurement is performed by the cadmium attachment / detachment method (see, for example, Non-Patent Document 1 and Patent Document 1), which attaches and detaches cadmium, a neutron absorber.

[0011] On the other hand, the second step requires the introduction and maintenance of new equipment, but the possibility of generating peculiar fuel is extremely low, so it is rarely used. Therefore, it is disadvantageous in terms of cost.

[0012] If it is possible to measure the effective neutron multiplication factor equivalent to the second step from the data of the burnup measurement device, which is the first step, it will be possible to accommodate unique fuel and maintain the functionality of the second step even if the equipment used in the conventional second step is eliminated.

[0013] As mentioned above, the neutron multiplication factor k is used as an indicator of subcriticality. k is the ratio of neutron production to neutron annihilation; k=1 indicates criticality, and k<1 indicates subcriticality. The unit dollar is sometimes used to express subcriticality. Using the delayed neutron production rate β, k / β is called a dollar. Even if the fuel actually becomes critical, explosive energy is not released immediately. When k exceeds β, it becomes an uncontrollable state called "prompt criticality." At this time, it is said that a reactivity of 1 dollar has been input. Sometimes cents are used as 1 / 100 of 1 dollar. When subcritical, k<1, but when k is small and far from 1, the subcriticality is called "deep," and when k is close to 1, the subcriticality is called "shallow."

[0014] Currently known subcriticality measurement techniques include: (1) Negative period method, (2) Control rod drop method, (3) Compensation method, (4) Neutron source multiplication method, (5) Inverse kinetics method, (6) Reactor noise analysis method, (7) Pulsed neutron source method, Among these methods, methods (1), (2), and (3) are used for reactors that go critical, and are not suitable for the fuel receiving process, which is always subcritical. Method (7) allows for absolute measurement of the degree of subcriticality without bringing the reactor to a critical state, but requires an accelerator to inject neutron pulses into the reactor, which requires new, large additional equipment and is not suitable for the purpose of removing the equipment in the second step. For these reasons, methods (4) to (6) are more suitable.

[0015] The neutron source multiplication method (4) can measure subcriticalities as deep as several tens of dollars, but requires calibration at a known subcritical state, which is difficult to achieve during the fuel receiving process, and is also not suitable as the second step.

[0016] The inverse dynamics method (5) estimates the degree of subcriticality from the change over time when the subcritical state fluctuates. During the fuel receiving process, measurements are performed in a static state, and there is no change in the degree of subcriticality, so this method is not suitable as the second step.

[0017] (6) Reactor noise analysis methods include (8) Rossi alpha method, (9) Feynman alpha method, (10) breakpoint frequency method, and (11) Mihalzozo method.

[0018] The Rossi α method in (8) is based on the assumption that the distribution of the time interval τ between neutron pulses observed in a nuclear reactor deviates from the Poisson distribution. -ατ Since the prompt neutron attenuation constant α has a component proportional to α={1-(1-β)k} / L ……(1) Therefore, the neutron multiplication factor is obtained by giving the delayed neutron production rate β and the prompt neutron lifetime L (see, for example, "Nuclear Reactor Physics Experiments" Corona Publishing).

[0019] The Feynman α method (9) utilizes the fact that the variance-to-mean ratio of the count value M of a neutron pulse within a certain time width deviates from “1” in the case of Poisson distribution to obtain the prompt neutron decay constant α in the same way as the Rossi α method (8) (see, for example, Non-Patent Document 2).

[0020] The break point frequency method (10) utilizes the fact that the frequency characteristics of the neutron count rate are determined by the reactor transfer function in the subcritical state, and that the break point frequency corresponds to the prompt neutron attenuation constant α.

[0021] Thus, all of the methods (8) to (10) calculate the prompt neutron attenuation constant α, and then input the constants β and L into equation (1) to obtain the neutron multiplication factor. In contrast, the Mihalzo method (11) allows for absolute measurement of the neutron multiplication factor, but requires the installation of a special neutron detector with a built-in neutron source Cf-252 in the reactor core and the target system, which requires the introduction of new equipment into the existing burnup acceptance process, and although the equipment itself is small, it is somewhat unsuitable for the purpose of removing the equipment in the second step.

[0022] All of the above technologies were developed for the purpose of measuring the neutron multiplication factor of a nuclear reactor. Patent Document 2 discloses a technology for monitoring criticality safety by applying the Rossi alpha method (8) or the Feynman alpha method (9) when storing spent fuel assemblies removed from a nuclear reactor in a storage rack or transport / storage container.

[0023] The delayed neutron fraction β and prompt neutron lifetime L are given in advance in the analysis. In the case of the spent nuclear fuel receiving process, the shape of the fuel assembly is fixed, and the ratio of water to nuclear fuel is also almost constant, so the changes in β and L are small, and the impact of errors caused by giving them in the analysis is thought to be small.

[0024] The above methods (8) to (10) can estimate the degree of subcriticality by processing time-series data from existing neutron detectors, so there is no need to introduce new equipment. Of the methods (8) to (10), only the Feynman α method (9) is considered applicable to deep subcriticality, and is a suitable method for replacing the second step.

[0025] Even when using the latest electronic measurement techniques, these reactor noise methods require measurement times of several minutes to several tens of minutes, and several hours to improve the measurement accuracy. However, since the second step is rarely used, the measurement time constraints are not thought to be a problem. [Prior art documents] [Patent documents]

[0026] [Patent Document 1] JP 2014-137259 A [Patent Document 2] JP 2005-106611 A [Non-patent literature]

[0027] [Non-Patent Document 1] Toshiba Corporation TLR-R001 Burnup measurement device for reprocessing facilities [Non-Patent Document 2] Kyoto University Reactor KUCA Experimental Text, "Neutron Correlation Experiments Feynman-α Method", www.rri.kyoto-u.ac.jp / CAD / Insei / 2003Text / Chap4_7‐2_r0.pdf. Summary of the Invention [Problem to be solved by the invention]

[0028] The burnup measurement equipment is equipped with a second step that can estimate the neutron multiplication factor to deal with unique fuels, but it is almost never used, which is one of the causes of rising costs in maintaining the equipment. If it is possible to estimate the neutron multiplication factor equivalent to the second step from the information on gamma rays and neutrons measured in the first step, it will be possible to eliminate the conventional second step equipment without compromising reliability and reduce costs.

[0029] The present invention has been made in consideration of the above-mentioned conventional circumstances, and aims to provide a subcriticality measurement device for spent nuclear fuel and a criticality management method for spent nuclear fuel that can reduce costs and enable estimation of the neutron multiplication factor, which is the function of the second step. [Means for solving the problem]

[0030] The subcriticality measuring device for spent nuclear fuel according to the embodiment includes: Sent to a reprocessing facility Spent Nuclear Fuel Neutrons emitted from A first step of measuring the burnup by measuring the neutron count of the spent nuclear fuel, and a second step of measuring the burnup of the spent nuclear fuel that does not meet the burnup limit in the first step. and, a time-series data analysis means for calculating a prompt neutron decay constant α from the time-series data based on a reactor noise method; a pre-analysis means for calculating a delayed neutron fraction β and a prompt neutron lifetime L in advance by neutron transport analysis; and a neutron multiplication factor calculation means for calculating a neutron multiplication factor k from the prompt neutron decay constant α, the delayed neutron fraction β, and the prompt neutron lifetime L, and the subcriticality of the second step is determined based on the neutron multiplication factor k. Effect of the Invention

[0031] According to the embodiment, it is possible to provide a subcriticality measurement device for spent nuclear fuel and a criticality management method for spent nuclear fuel that can reduce costs and enable estimation of the neutron multiplication factor, which is the function of the second step. [Brief description of the drawings]

[0032] [Figure 1] FIG. 1 is a diagram showing a schematic configuration of a subcriticality measurement device for spent nuclear fuel and a criticality management method for spent nuclear fuel according to an embodiment. [Diagram 2] FIG. 2 is a flow diagram of the subcriticality measurement device for spent nuclear fuel and the criticality management method for spent nuclear fuel according to the embodiment. [Diagram 3] Graph showing an example of experimental results obtained in the critical experiment facility. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0033] Hereinafter, the subcriticality measurement device for spent nuclear fuel and the criticality management method for spent nuclear fuel according to the embodiment will be described with reference to the drawings. The subcriticality measurement device for spent nuclear fuel and the criticality management method for spent nuclear fuel according to the embodiment are applied to a reprocessing facility when accepting transported spent nuclear fuel.

[0034] FIG. 1 is a schematic diagram showing the general configuration of a subcriticality measurement device for spent nuclear fuel and a criticality management method for spent nuclear fuel according to an embodiment, and FIG. 2 shows a flow of a second step substitution by a first step signal from a burnup measurement device.

[0035] As shown in Fig. 1, two neutron detectors, neutron detector 1 and neutron detector 2, for measuring neutrons from the fuel assembly 3 are provided around a spent fuel assembly 3. As shown in Fig. 1, this embodiment includes a first step execution unit 100 and a second step execution unit 200. If the burnup limit value is not satisfied in the first step performed by the first step execution unit 100, the second step is executed by the second step execution unit 200.

[0036] As shown in FIG. 1, the second step execution unit 200 includes a neutron count storage unit 201 , a time-series data analysis unit 202 , a pre-analysis unit 203 , and a neutron multiplication factor calculation unit 204 .

[0037] Neutron counting data from neutron detector 1 and neutron detector 2 is input to a first step execution unit 100 and is stored in a neutron count storage unit 201 as time-series data.

[0038] As shown in FIG. 2, in this embodiment, neutron counting data is acquired (step 301 in FIG. 2), and in the first step of measuring the burnup of the spent nuclear fuel (step 302 in FIG. 2), for spent nuclear fuel that does not satisfy the burnup limit value and has an indication value exceeding the burnup limit value (No in step 303 in FIG. 2), the time series data analysis unit 202 of the second step execution unit 200 is activated to perform the second step measurement (step 304 in FIG. 2).

[0039] In order to execute the first step, the time-series data analysis unit 202 of the second step execution unit 200 uses the neutron counting data output from the neutron detector 1 and the neutron detector 2, which are accumulated in the neutron counting accumulation unit 201. The time-series data is accumulated for several minutes to several tens of minutes and analyzed in the time-series data analysis unit 202, and a Feynman alpha analysis is performed to obtain the prompt neutron decay constant α.

[0040] A specific processing method of the Feynman α method will be described below. The frequency of radioactive decay is based on the Poisson distribution, but it is known that it deviates from the Poisson distribution when multiple neutrons are generated simultaneously or when a chain reaction occurs due to fissile material (see, for example, Non-Patent Document 2). The ratio of the variance to the mean (variance-to-mean ratio) is used as an index of deviation from the Poisson distribution. Over a certain time interval t i ~t i Assuming that the neutron count at +Δt is M(i) and there are imax pieces of data, the average, variance, and variance / average ratio can be expressed as follows: The time-series data analysis unit 202 of the second step execution unit 200 sequentially calculates the average (step 305 in FIG. 2), the variance (step 306 in FIG. 2), the variance-to-average ratio (step 307 in FIG. 2), and the prompt neutron decay constant α (step 308 in FIG. 2), as shown below.

[0041] average

number

[0042] dispersion

number

[0043] variance / mean ratio

number

[0044] Y=C[1-{1-exp(-αΔt)} / αΔt] where C is a constant.

[0045] To calculate the prompt neutron attenuation constant α, Y is plotted as Δt is changed, and the prompt neutron attenuation constant α is obtained by fitting with the above function. The graph in Figure 3 shows the results of an example experiment obtained at the Toshiba Critical Assembly (NCA), and it can be seen that the prompt neutron attenuation constant α can be obtained from the relationship between the gate width (Δt) and Y.

[0046] The water level shown in the figure indicates the water level (mm) of the reactor core tank of the critical experiment facility. The higher the water level, the higher the neutron multiplication factor, and a water level of 1000 is closer to the critical state than a water level of 900. The gate width corresponds to the above-mentioned Δt. The constant C can also be found at the same time by fitting.

[0047] Furthermore, the pre-analysis unit 203 obtains the delayed neutron fraction β and the prompt neutron lifetime L in advance by neutron transport analysis. Then, the neutron multiplication factor calculation unit 204 obtains the neutron multiplication factor k using the delayed neutron fraction β, the prompt neutron lifetime L, and the prompt neutron attenuation constant α according to the above-mentioned relational expression (1) shown below (step 309 in FIG. 2). α={1-(1-β)k} / L …(1)

[0048] This method does not depend on the absolute value of the neutron count rate, and therefore has the advantage that it can estimate the neutron multiplication factor k without being affected by fluctuations in the neutron count rate, even for fuels that have a unique relationship between burnup and neutron intensity or gamma ray intensity due to the influence of operating history, etc.

[0049] If the neutron multiplication factor k thus obtained falls below a set value, for example k=0.8 (Yes in step 310 in FIG. 2), no abnormality is detected and the process is carried out in the normal route (step 311 in FIG. 2). On the other hand, if the neutron multiplication factor k exceeds the set value (No in step 310 in FIG. 2), an abnormality is detected and measures such as isolation are taken (step 312 in FIG. 2).

[0050] As described above, according to the embodiment of the subcriticality measurement device for spent nuclear fuel and the criticality management method for spent nuclear fuel, it is possible to reduce costs and enable estimation of the neutron multiplication factor, which is the function of the second step.

[0051] In Figure 1, the time series data is analyzed using the Feynman alpha method, but the Rossi alpha method or the breakpoint frequency method may also be used. If a special neutron detector, such as a Cf-252 neutron source, is available, the Mihalzo method may also be used. In addition, multiple methods among these may also be used.

[0052] By analyzing multiple pieces of information, the cause of the anomalous fuel can be identified early on, and even if a decision is made to isolate it, it will be possible to determine how the anomalous fuel came about.

[0053] The analysis of time-series data may be constantly performed as a backup and not normally used to determine whether or not there is an abnormality in the process, but may be used to determine whether or not there is an abnormality only when peculiar fuel occurs.

[0054] The most suitable neutron transport analysis method is the continuous energy Monte Carlo method, but the Sn transport calculation method or the diffusion calculation method may also be used.

[0055] Continuous energy Monte Carlo methods include MCNP, developed at the Los Alamos National Laboratory in the United States, and MVP, developed at the Japan Atomic Energy Agency. Sn transport calculation methods include DOORS, developed at the Oak Ridge National Laboratory in the United States. Diffusion calculation codes include CITATION, developed at the Oak Ridge National Laboratory in the United States (for example, see the "Radiation Shielding Handbook (Basics)" by the Atomic Energy Society of Japan).

[0056] In the Feynman alpha method, the covariance of detector 1 and detector 2 may be used instead of the variance. Reactor noise analysis using the covariance is also shown in literature such as "Measurement of Effective Delayed Neutron Fraction Based on the Covariance Method" in the Journal of the Atomic Energy Society of Japan, vol. 37, No. 6 (1995).

[0057] Although several embodiments of the present invention have been described above, these embodiments are presented as examples and are not intended to limit the scope of the invention. These novel embodiments can be implemented in various other forms, and various omissions, substitutions, and modifications can be made without departing from the spirit of the invention. These embodiments and their modifications are included in the scope and spirit of the invention, and are included in the scope of the invention and its equivalents described in the claims. [Explanation of symbols]

[0058] 1...Neutron detector, 2...Neutron detector, 3...Fuel assembly.

Claims

1. A first step of measuring burnup by measuring the neutron count of neutrons generated from spent nuclear fuel sent to a reprocessing facility; A subcriticality measurement device for spent nuclear fuel, which performs a second step of measurement on the spent nuclear fuel that does not satisfy a burnup limit value in the first step, A storage means for storing the neutron counts as time-series data; A time series data analysis means for calculating a prompt neutron decay constant α from the time series data based on a reactor noise method; A pre-analysis means for calculating a delayed neutron fraction β and a prompt neutron lifetime L in advance by a neutron transport analysis; a neutron multiplication factor calculation means for calculating a neutron multiplication factor k from the prompt neutron attenuation constant α, the delayed neutron fraction β, and the prompt neutron lifetime L; Equipped with The subcriticality of the second step is determined based on the neutron multiplication factor k. A device for measuring subcriticality of spent nuclear fuel.

2. The time series data analysis means extracts from imax pieces of data of neutron counts M(i) in the time series data during a time interval from ti to ti+Δt: [0010] The average is calculated by [0025] The variance is calculated by [0030] The variance to mean ratio is calculated by Plot Y when the Δt is changed and compare it with the constant C. Y=C[1-{1-exp(-αΔt)} / αΔt] The prompt neutron attenuation constant α is calculated by fitting 2. The apparatus for measuring subcriticality of spent nuclear fuel according to claim 1.

3. The time series data analysis means calculates the covariance of the two neutron detectors instead of calculating the variance.

3. The subcriticality measuring device for spent nuclear fuel according to claim 2.

4. A first step of measuring burnup by measuring neutron counting of neutrons generated from spent nuclear fuel sent to a reprocessing facility; A method for managing criticality of spent nuclear fuel, comprising: performing a second step of measurement on the spent nuclear fuel that does not satisfy a burnup limit value in the first step, accumulating the neutron counts as time-series data; A time series data analysis step of calculating a prompt neutron decay constant α from the time series data based on a reactor noise method; a pre-analysis step of calculating a delayed neutron fraction β and a prompt neutron lifetime L in advance by neutron transport analysis; a neutron multiplication factor calculation step of calculating a neutron multiplication factor k from the prompt neutron attenuation constant α, the delayed neutron fraction β, and the prompt neutron lifetime L; Equipped with The subcriticality of the second step is determined based on the neutron multiplication factor k. A method for managing criticality of spent nuclear fuel.

5. The time series data analysis step includes: In the time series data, the neutron count M(i) in the time interval ti to ti+Δt From imax pieces of data, [0045] The average is calculated by [0050] The variance is calculated by [006] The variance to mean ratio is calculated by Plot Y when the Δt is changed and compare it with the constant C. Y=C[1-{1-exp(-αΔt)} / αΔt] The prompt neutron attenuation constant α is calculated by fitting 5. The method for managing criticality of spent nuclear fuel according to claim 4.

6. 6. The method for managing criticality of spent nuclear fuel according to claim 5, wherein in said time series data analysis step, instead of calculating said variance, a covariance of two neutron detectors is calculated.

7. 7. The method for criticality management of spent nuclear fuel according to claim 4, further comprising isolating the spent nuclear fuel when the neutron multiplication factor k exceeds a preset value.

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