JP4755061B2 - Nuclear facility leakage monitoring system and leakage monitoring method thereof - Google Patents

Description

  The present invention relates to a leakage monitoring system for a nuclear facility and a leakage monitoring method thereof, and more particularly, to a leakage monitoring system for a nuclear facility suitable for application to a building of a boiling water nuclear power plant and a leakage monitoring method thereof.

  As a leakage water monitoring system with the current nuclear power plan, a leakage detection system (Leak Detecting System, hereinafter referred to as LDS) in the reactor containment vessel is used. In a boiling water nuclear power plant, a reactor pressure vessel (RPV) containing a reactor core is installed in a reactor containment vessel installed in a reactor building. This boiling water nuclear power plant has a recirculation system that is connected to a reactor pressure vessel and supplies cooling water to the reactor core, and a turbine, condenser, and feed water pump are installed in the turbine building. is doing. The recirculation system has a recirculation pipe provided with a recirculation pump. The steam generated in the core is sent to the turbine through the main steam pipe, condensed in the condenser, and returned to the water. This water is boosted by the feed pump and returned to the reactor pressure vessel.

  In the conventional LDS, the gas in the reactor containment vessel is collected by driving the blower from a plurality of sampling ports provided in the reactor containment vessel, and the collected gas is collected at the radioactivity measurement unit outside the reactor containment vessel. Measure the radioactivity of In the measurement of radioactivity, the trapped gas from a plurality of sampling ports is guided to a roll filter, and the dust contained in the trapped gas is collected by the filter. The amount of radioactivity contained in the solid radioactive material contained in the captured dust and the radioactive gas contained in the repair gas is measured by a scintillation detector. In this way, the amount of radioactivity in the reactor containment vessel is continuously measured, and the presence or absence of leaked water in the reactor containment vessel is monitored.

  Leakage monitoring in the turbine building of a boiling water nuclear power plant is provided with gas sampling ports in each room in the turbine building, and gas is collected selectively from these sampling ports, and the collected gas is collected. This is done by measuring the amount of radioactivity.

  In nuclear facilities such as boiling water nuclear power plants, in order to ensure plant safety when cracks occur in piping, etc. and cooling water exceeding a set amount leaks in at least one of gas and liquid states Shut down the reactor. In addition, after shutting down the reactor, it is necessary to identify the leak location and take necessary measures. As a method for identifying a leak location, a method has already been proposed in which it is determined whether a leak has occurred in a reactor coolant system or a steam system.

  The determination method is described in JP-A-59-150388, JP-A-2-159599, and JP-A-5-249278. JP-A-59-150388 discloses a hydrogen water concentration in a reactor containment vessel and a radionuclide radioactivity concentration in a boiling water nuclear power plant, and a radioactivity in leaked water in the reactor containment vessel. The concentration is measured, and using these measured values, it is determined whether the leak point is in the steam system or the reactor water system. As the radionuclide radioactivity concentration, radioactivity concentrations of I-131 and Na-24 are measured. Japanese Patent Application Laid-Open No. 2-159599 determines whether a steam system or a reactor water system when a leak occurs in a reactor containment vessel of a boiling water nuclear power plant. This determination is performed by measuring the concentration and composition of the radionuclide in the atmosphere in the reactor containment vessel, and comparing the measurement result with the concentration ratio of the radionuclide contained in the reactor water and main steam, respectively. As the radionuclide concentration ratio, N-13 / F-18 concentration ratio or I-131 / I-133 concentration ratio is used. According to the leakage monitoring method described in Japanese Patent Laid-Open No. 5-249278, the hydrogen concentration in the atmosphere in the reactor containment vessel is measured, and the amount of condensed water drain generated from the dehumidification system in the reactor containment vessel or the reactor containment vessel Based on the ratio of the measured hydrogen concentration to the amount of drainage of the inner sump, it is determined whether the leaked portion is a steam system or a reactor water system.

JP 59-150388 A Japanese Patent Laid-Open No. 2-159599 JP-A-5-249278

  The leakage monitoring methods for nuclear facilities described in the above-mentioned JP-A-59-150388, JP-A-2-159599, and JP-A-5-249278 have the following problems. Japanese Patent Application Laid-Open No. 59-150388 uses the radioactivity concentrations of I-131 and Na-24, which are radionuclides, to determine whether the leaked portion is a steam system or a reactor water system. However, I-131 is a fission product and cannot be detected with high sensitivity unless the fuel rods loaded in the reactor core are damaged. Na-24 is generated when seawater leaks in the heat transfer tube of the condenser and Na flows into the reactor core. The probability of occurrence of fuel rod breakage in the core and seawater leaks in the heat transfer tubes of the condenser is extremely small in current reactors, and those accidents can hardly occur. For this reason, it is difficult to determine the leak location based on the radioactivity concentrations of I-131 and Na-24.

  Japanese Patent Application Laid-Open No. 2-159599 shows a decay curve of the radioactivity amount of the radionuclide released into the reactor containment vessel in FIG. According to FIG. 2, the decay curve of N-13 activity is represented by curve a, and the decay curve of F-18 activity is represented by curve b. In addition, when the leakage location is a reactor water system, the attenuation curve of the sample is a curve c, and when the leakage location is a steam system, the attenuation curve of the sample is a curve c. The decay curve of these samples is a composite curve of curve a and curve b and depends on the concentration ratio of N-13 and F-18. However, as is clear from FIG. 2, the location of leakage by N-13 and F-18 cannot be determined unless the radioactivity concentration is measured for 30 minutes to 1 hour after the occurrence of a leakage accident. The responsiveness until judgment is bad. Further, since I-131 and I-133 are fission products, it is not appropriate to use them as indicators for determining a leaked portion, as in JP-A-59-150388. Japanese Patent Application Laid-Open No. 5-249278 also requires that the hydrogen concentration be measured over a long period of time after the occurrence of a leakage accident, resulting in poor response to judgment.

  As described above, it is difficult for conventional leak monitoring systems to determine whether a leak is a steam system or a reactor water system, or it takes a long time to determine the propagation of a leaked water accident and start of leak repair work. It has been found that there is a problem to delay.

  An object of the present invention is to provide a leakage monitoring system for a nuclear facility and a leakage monitoring method thereof capable of accurately detecting a leakage location in a short time.

  The feature of the present invention that achieves the above object is to detect radiation from radionuclides contained in gas sampled from within a nuclear facility equipped with steam system piping and reactor cooling water system piping, and the radiation obtained by this detection. Based on the radioactivity of at least one of radioactive nitrogen and radioactive carbon and the radioactivity of the corrosion product contained in the nuclide analysis information obtained by the nuclide analysis using the detection signal It is to determine whether it is in the steam system piping and in the reactor cooling water system piping.

  In the present invention, the position of the leakage point is determined on the basis of the radioactivity of at least one of radioactive nitrogen and radiocarbon and the radioactivity of the corrosion product obtained by nuclide analysis. It can be accurately determined in a short time which of the water-based piping is causing the leakage.

  Preferably, the radioactivity amount of the first radionuclide selected from N-13, N-16 and C-15 having different half-lives contained in the nuclide analysis information and the first radionuclide selected from them. It is desirable to specify the position of the leaking location occurring in the steam system piping based on the ratio of the radioactivity of the two radionuclides.

  The location of the leak location occurring in this steam system piping was made based on the new knowledge found by the inventors described below. As a result of various studies, the inventors have determined that the ratio of the N-13 radioactivity to the N-16 radioactivity, ie, (N-13 radioactivity) / (N-16 radioactivity) (hereinafter, referred to as “N-16 radioactivity”). (N-13 / N-16 ratio), a new finding has been found that a leak point in a steam system can be accurately estimated. This new knowledge will be explained specifically.

  The γ-ray energy of the radionuclide contained in the cooling water of the reactor cooling water system is 834 KeV for Mn-54 and 1.25 MeV for Co-60. The energy of the radionuclide contained in the vapor of the vapor system is 511 keV for N-13 and 6.1 MeV for N-16. These radionuclides can be analyzed by using a semiconductor radiation detector capable of detecting even high-energy γ-rays. The half-lives of N-13 and N-16 are greatly different, with the former being 10 minutes and the latter being 7.1 seconds. The N-13 / N-16 ratio in the reactor is constant in the steady state operation of the reactor. The N-13 / N-16 ratio depends on the time difference to the assumed leakage points (for example, flanges and joints) of various instrumentation pipes connected to the main steam pipe via the main steam pipe connected to the reactor. It changes at a predetermined rate. That is, the position of the leaked portion in the steam system can be accurately estimated by obtaining the main components of the leaked gas and the radioactivity amounts of these main components.

  Instead of the N-13 / N-16 ratio, the ratio of the N-16 radioactivity to the C-15 radioactivity, ie, (N-16 radioactivity) / (C-15 radioactivity) (hereinafter N -16 / C-15 ratio) or the ratio of N-13 radioactivity to C-15 radioactivity, ie, (N-13 radioactivity) / (C-15 radioactivity) (hereinafter N-13) / C-15 ratio) can also be used to accurately estimate the leak location in the steam system.

  ADVANTAGE OF THE INVENTION According to this invention, it can be calculated | required with sufficient precision in a short time whether leakage has generate | occur | produced in either steam system piping or reactor cooling water system piping. For this reason, a leak location can be known quickly, and maintenance monitoring and repair work for a nuclear power plant can be quickly prepared. The safety of nuclear facilities is significantly improved.

  Embodiments of the present invention will be described below with reference to the drawings.

  Before describing the leak detection system (Leak Detecting System, hereinafter referred to as LDS) of the nuclear facility of the first embodiment which is a preferred embodiment of the present invention, a nuclear facility to which LDS is applied, for example, boiling water nuclear power An outline of the power plant will be described with reference to FIG. A boiling water nuclear power plant has a reactor containment vessel 1 installed in a reactor building (not shown), and a reactor pressure vessel (hereinafter referred to as RPV) 4 containing a core 3 is contained in the reactor. It is installed in the container 1. The recirculation system has a recirculation pipe 5 connected to the RPV 4 and a recirculation pump 6 attached to the recirculation pipe 6. This recirculation system is also installed in the reactor containment vessel 1. The boiling water nuclear power plant has a turbine building 2, and a turbine 7 and a condenser 8 are installed in the turbine building 2. The turbine 7 is connected to the RPV 4 by the main steam pipe 10. The condenser 8 is connected to the RPV 4 by a water supply pipe 11. A water supply pump 9 is provided in the water supply pipe 11.

  The cooling water boosted by the recirculation pump 6 is supplied to a jet pump (not shown) installed in the RPV 4 through the recirculation pipe 5. Cooling water discharged from the jet pump is supplied to the core 3 and heated in the core 3. Part of the heated cooling water becomes steam. This steam is supplied to the turbine 7 through the main steam pipe 10 to drive the turbine 7. A generator (not shown) connected to the turbine 7 rotates to generate electric power. The steam discharged from the turbine 7 is condensed by heat exchange with the seawater supplied by the cooling water pipe 53 in the condenser 7 to become water. This water is boosted by a feed water pump 9 or the like as feed water and supplied to the RPV 4 through a feed water pipe 11.

  The LDS 20 of this embodiment will be described with reference to FIGS. 1 and 2 below. The LDS 20 includes a plurality of sampling ports 13 arranged at predetermined positions in the reactor containment vessel 1, a plurality of sampling ports 21a, 21b, 21c, 21d, 21e,. Measuring sections 30A and 30B, wave height analyzers (nuclide analyzers) 34A and 34B, a data processing device (leakage location specifying device) 36, a storage device 35, an operation panel 37, and a pipe line 23 provided with a blower 12 are provided. Each sampling pipe 49 having a plurality of sampling ports 13 respectively arranged at predetermined positions in the reactor containment vessel 1, that is, sampling ports 13a, 13b, 13c, 13d, 13e,... (See FIG. 2). Is connected to the pipe line 23 via the selector valve 14A. The sampling ports 13a, 13b, 13c, 13d, 13e,... Are pressure detection systems that are steam systems (for example, main steam pipe 10 and branch pipes connected to the main steam pipe 10) located in the reactor containment vessel 1. Instrumentation pipes such as pipes) and reactor water systems such as reactor cooling water systems (for example, recirculation systems, reactor purification systems, residual heat removal systems, water supply pipes, and instrumentation pipes that are branch pipes thereof) Placed near each other. Most of the pipe line 23 is disposed outside the reactor containment vessel 1, and both ends thereof are disposed within the reactor containment vessel 1. A blower 12 is provided in the conduit 23. A radioactivity measurement unit 30 </ b> A is provided in the pipeline 23 outside the reactor containment vessel 1. A drain water receiving portion 17 is formed at the bottom of the dry well in the reactor containment vessel 1, and a drain water meter 38 is installed in the drain water receiving portion 17. A dew point meter 18 and a hydrogen concentration meter (not shown) are installed in the reactor containment vessel 1. Each sampling pipe 52 individually having sampling ports 21a, 21b, 21c, 21d, 21e,... Arranged at a predetermined position in the turbine building 2 (for example, each room in the turbine building 2) is a selector valve. 14B is connected to the pipe line 28. Sampling ports 21 a, 21 b, 21 c, 21 d, 21 e,... Are steam systems (for example, main steam pipe 10, pressure detection pipes that are branch pipes connected to main steam pipe 10, etc.) located in turbine building 2. Instrumentation piping) and a reactor cooling water system (for example, a water supply pipe and each instrumentation pipe that is a branch pipe thereof). The pipe line 28 is arranged in the turbine building 2, most of the pipe line 28 is arranged in a predetermined room (for example, a monitor room) in the turbine building 2, and both ends are arranged in each room in the turbine building 2. A radioactivity measurement unit 30 </ b> B is provided in the pipeline 28 in a predetermined room in the turbine building 2. A blower 29 is provided in the conduit 28. The radioactivity measurement unit 30 </ b> B can be disposed outside the turbine building 2.

  The detailed structure of the radioactivity measurement unit 30A will be described with reference to FIG. The radioactivity measurement unit 30A includes a roll filter (filter device) 15A, a Ge radiation detector (hereinafter referred to as Ge detector) 16A that is a semiconductor radiation detector, and a measurement chamber 24A. The measurement chamber 24A is attached to the pipe line 23. The roll filter 15A and the Ge detector 16A are attached to the measurement chamber 24A. For the Ge detector 16A and a Ge detector 16B described later, a Ge detector having a relative efficiency of 40% or more that can easily detect high-energy γ rays is used. An amplifier 33A is connected to the Ge detector 16A. The roll-shaped filter 15A has a pair of rollers 25 and 26 and a band-shaped filter 27 that are rotating bodies. The belt-like filter 27 is wound around the rollers 25 and 26, respectively. The belt-like filter 27 is wound around the roller 26 from the other roller 25 by intermittently rotating one roller, for example, the roller 26 at predetermined time intervals. The radioactivity measurement unit 30B also has the same configuration as the radioactivity measurement unit 30A. However, the roll filter, Ge detector, amplifier, and measurement chamber provided in the radioactivity measurement unit 30B are represented by the roll filter 15B, Ge detector 16B, amplifier 33B, and measurement chamber 24B. As the semiconductor radiation detector, each semiconductor radiation detector using CdTe and CZT separately may be used instead of the Ge detector.

  The operation panel 37 includes an arithmetic device 37A, a display device 37B, and a control device 37C. The display device 37B is connected to the arithmetic device 37A. The control device 37C is connected to the selector valves 14A and 14B.

  The amplifier 33A of the radioactivity measurement unit 30A is connected to the wave height analyzer 34A by a wiring 54. The amplifier 33B of the radioactivity measurement unit 30B is connected to the wave height analyzer 34B by a wiring 38. The wave height analyzers 34A and 34B are connected to the data processing device 36. A storage device 35 is connected to the data processing device 36. The data processing device 36 is connected to the arithmetic device 37A. The dew point meter 18 is connected to the arithmetic unit 37A by a wiring 47. The drain water meter 38 is connected to the arithmetic unit 37A by a wiring 39. A hydrogen concentration meter (not shown) installed in the reactor containment vessel 1 is also connected to the arithmetic unit 37A.

  The basic concept of the estimation of the position of the leaked part in the steam system using the above-described N-13 / N-16 ratio newly found by the inventors will be described in detail below. A radiation detection signal output from a semiconductor radiation detector (for example, Ge detector 16A) is input to a wave height analyzer and subjected to nuclide analysis. An example of a nuclide analysis measurement spectrum (nuclide analysis information) obtained by this nuclide analysis is shown in FIG. Since the maximum energy of the measurement target nuclide is 6.1 MeV of N-16, the semiconductor radiation detector needs a measurement energy range of about 8 MeV at the maximum. Moreover, since N-13 is 511 KeV and Co-60 is 1.25 MeV, the semiconductor radiation detector needs a measurement energy range of 0 KeV or more.

  FIG. 4 shows decay curves of N-13 (half-life: 10 minutes) and N-16 (half-life: 7 seconds) shown in FIG. The amount of N-13 and N-16 radioactivity present in the RPV 4 during steady operation of the nuclear power plant is balanced by the generation loss due to the nuclear reaction and the transition loss from the RPV 4 to the steam discharged to the main steam pipe 10. It is kept at a certain value. The respective radioactivity amounts of N-13 and N-16 existing in the RPV 4 during the steady operation are measured by routine analysis work of facility management. These radioactivity levels can be collected as needed. The respective nucleation reactions of N-13 and N-16 are O-16 (p, 2n) N-13 and O-16 (n, p) N-16. When N-13 and N-16 have just moved from the liquid phase region (reactor water) to the gas phase region (steam generated in the core 3) in the RPV 4, there is an N-13 / N-16 ratio of steam. Held in value. However, the N-13 / N-16 ratio, which is the ratio of radioactivity, varies depending on the elapsed time after the transition of N-13 and N-16 to steam (see FIG. 4). This ratio depends only on the half-life of both radionuclides. The decay rate of N-16 having a short half-life is significantly greater than that of N-13.

  For various values of the N-13 / N-16 ratio, the inventors change how these values vary with the elapsed time after the transition to N-13 and N-16 steam. We examined whether. FIG. 5 shows the result of the study, and the N-13 / N-16 ratio value (indicated by F) at the time when N-13 and N-16 have just moved from the liquid phase region to the gas phase region in RPV4. ) As a parameter, the change in the N-13 / N-16 ratio with respect to the elapsed time is shown. Four types of F, 1000, 100, 10 and 1, were used. Even if the value of F, which is the initial value, changes, the slope of the decrease in the N-13 / N-16 ratio with respect to the elapsed time is the same. Based on the results shown in FIG. 5, the inventors have found that N-13 and N-16 reach the place where there is a possibility of vapor leakage (for example, the place where equipment and parts are installed) from RPV4. When the time required (arrival time) is known, a new finding has been found that the N-13 / N-16 ratio at a place where there is a possibility of steam leakage can be understood. The arrival time corresponds to the elapsed time. According to this knowledge, if the arrival time is known, the leakage in the steam system using the N-13 / N-16 ratio calculated by using the radioactivity amounts of N-13 and N-16 obtained by nuclide analysis. The position of the location can be estimated.

  In the core, C-15, which is a radionuclide, is generated by the nuclear reaction of O-16. Similarly to N-13 and N-16, this C-15 can also be used for estimation of the position of the leak location in the steam system. The N-16 / C-15 ratio or N-13 / C-15 ratio obtained using the radioactivity amount of C-15 obtained by nuclide analysis is also different in slope, but N-13 / N shown in FIG. Properties similar to the -16 ratio can be obtained. For this reason, even if it uses N-16 / C-15 ratio or N-13 / C-15 ratio, the position of the leak location of a steam system can be estimated.

  A branch pipe 50 connected to the main steam pipe 10 and an instrumentation pipe 51 connected to the branch pipe 50 are schematically shown in FIG. The instrumentation pipe 51 is, for example, an instrumentation pipe for measuring the pressure of steam. PF-1 is a flange for connecting the branch pipe 50, PT-1 is a joint for an instrument, and PT-2 to PT-6 are pipe joints for the instrument pipe 51. PV-1 is a valve provided in the instrument pipe 51. The installation location of each component (flange, joint, valve, etc.) provided in the branch pipe 50 and the instrumentation pipe 51 can be grasped in advance from the design book, and the design of the branch pipe 50 and the instrumentation pipe 51 is designed. The flow rate can also be grasped in advance. The arrival times of N-13 and N-16 from the core to each component can also be grasped from the design document. Based on these pieces of information, it is possible to grasp the arrival time of N-13 and N-16 and the N-13 / N-16 ratio up to each component.

  Based on the information such as the above-mentioned design books, a database was created in which information on each component installation location (location where leakage could occur) and N-13 / N-16 ratio information was related. An example of this database information is shown in FIG. Although the arrival time is shown in FIG. 7, the created database information does not include the arrival time. This database information is stored in the storage device 35 in advance. Using the N-16 / C-15 ratio or N-13 / C-15 ratio, the location of each component, and the arrival time, database information similar to the N-13 / N-16 ratio is created, and a storage device 35 may be stored.

  The leakage monitoring using the LDS 20 will be specifically described. A control device 37C provided on the operation panel 37 controls the selector valve 14A to sequentially connect the sampling ports 13a, 13b, 13c, 13d, 13e,. Since the blower 12 is driven, the gas sampled from each sampling port at a predetermined position in the reactor containment vessel 1 is sequentially supplied to the pipe line 23 via the corresponding sampling pipe 49. This gas is introduced into the measurement chamber 24A through the conduit 23. The Ge detector 16A emits gamma rays emitted from radionuclides (for example, N-13, N-16, C-15, Mn-54, Co-60) contained in the gas that has reached the measurement chamber 24A. To detect. The Ge detector 16A outputs a γ-ray detection signal, and this γ-ray detection signal is amplified by the amplifier 33A. A steam system connected to the RPV 4 in the reactor containment vessel 1 (for example, an instrumentation pipe such as a main steam pipe 10 and a pressure detection pipe which is a branch pipe connected to the main steam pipe 10) and an atom which is a reactor water system When there is no leakage of steam or cooling water from the reactor cooling water system (for example, recirculation system, reactor purification system, residual heat removal system, feed water pipe, and instrumentation pipes that are branch pipes of these) The Ge detector 16A outputs a 0V γ-ray detection signal. It is assumed that a crack generated in the main steam pipe 10 passes through the reactor containment vessel 1 and steam leaks from the main steam pipe 10 into the reactor containment vessel 1. At this time, the gas introduced into the measurement chamber 24A contains a radionuclide (for example, N-13, N-16, C-15). The Ge detector 16A detects γ rays emitted from the radionuclide or the like. It is assumed that a crack generated in the reactor purification system pipe connected to the RPV 4 penetrates and the cooling water evaporates and leaks from the reactor purification system pipe into the reactor containment vessel 1. At this time, the gas introduced into the measurement chamber 24A contains solid radionuclides (for example, Mn-54, Co-60), and the Ge detector 16A emits gamma rays emitted from the solid radionuclides and the like. Is detected. Solid radionuclide (for example, Mn-54, Co-60) is trapped by the band filter 27 when the gas passes through the band filter 27 in the measurement chamber 24A. The Ge detector 16 </ b> A detects γ-rays from the solid radionuclide captured by the band-like filter 27. The band-shaped filter 27 is moved intermittently in the measurement chamber 24A as described above.

  The γ-ray detection signal output from the amplifier 33A is transmitted to the wave height analyzer 34A. The wave height analyzer 34A performs online nuclide analysis based on the input γ-ray detection signal. Information obtained by the nuclide analysis (see FIG. 3) is input to the data processor 36. The amplifier 33A also outputs information on the total count value per unit time (hereinafter referred to as radioactivity measurement value) based on the input γ-ray detection signal to the data processing device 36. The data processing device 36 executes processing according to the processing procedure shown in FIG. 8 using the input nuclide analysis information and the information of the database previously created and stored in the storage device 35.

  First, using the input nuclide analysis information, it is determined whether or not the amount of radioactivity has increased for each radionuclide (step 61). If there is a radionuclide having an increased amount of radioactivity, the determination in step 61 is “Yes”. If the nuclide does not exist, “No” is determined. In the case of “Yes”, it is determined whether the amount of radioactivity exceeds a reference value (step 62). The determination in step 62 is “Yes” if the radioactivity amount of the corresponding radionuclide exceeds the reference value, and “No” if it does not exceed the reference value. If the determination in step 61 or 62 is “No”, message information of “no leakage” is created (step 67), and the process of step 61 is executed. In step 67, for example, message information “No leakage in reactor cooling water system and steam system” is created. When the determination in step 62 is “Yes”, it is determined whether the radionuclide whose radioactivity exceeds the reference value is the main component (CP) contained in the primary cooling water (reactor water) or a gas component (step) 63). When the amount of radionuclide (such as Co-60) contained in the primary cooling water exceeds the reference value, it is determined that the primary cooling water is leaked (step 65). In this case, the message information “A leak has occurred in the reactor cooling system” is created. In step 63, when the radionuclide of the gas component exceeds the reference value, it is determined that the steam is leaking, and the position of the leaking portion in the steam system is estimated (step 64). In step 64, an N-13 / N-16 ratio is calculated based on the nuclide analysis information. Using the obtained N-13 / N-16 ratio, the position of the corresponding leak location (for example, PF-1, PT-2) from the database information (for example, see FIG. 7) stored in the storage device 35. Etc.)). Information indicating the position of the leakage location estimated in the steam system is created (step 66). In step 66, for example, message information related to the leakage of the steam system “leakage occurred in PT-2 of the steam system” is created. Even if the N-16 / C-15 ratio or the N-13 / C-15 ratio is used instead of the N-13 / N-16 ratio in the step 64, the position of the leakage point in the steam system can be estimated. .

  The data processing device 36 outputs information created by the processing procedure of FIG. 8 and information such as nuclide analysis information and radioactivity measurement values input from the wave height analyzer 33A to the arithmetic device 37A of the operation panel 37. The measured values of the dew point meter 18, the drain water meter 38, and the hydrogen concentration meter are also input to the arithmetic unit 37A.

  When a leak occurs in the reactor containment vessel 1, the leaked water from the piping or the like flows into the drain water receiving portion 17. Steam that has leaked from a pipe such as the main steam pipe 10 is also collected in the drain receiving portion 17 when it is condensed into water in the dry well of the reactor containment vessel 1. The drain water amount is measured by a drain water meter 38 provided in the drain receiving portion 17. When the amount of drain water, that is, the amount of leaked water becomes, for example, 1 G / min (3.8 L / min) or more, the reactor is stopped by the scram. However, it is not possible to determine whether the leaked water is radioactive or non-radioactive only by measuring the amount of drain water. The hydrogen concentration meter and the dew point meter 18 installed in the reactor containment vessel 1 also obtain auxiliary data for monitoring the leakage of primary cooling water (reactor water) and steam into the reactor containment vessel 1.

  When the measured values of the dew point meter 18 and the drain water meter 38 rise while the message information “No leakage in the reactor cooling water system and steam system” is input, the arithmetic unit 37A It is determined that leakage has occurred in an auxiliary cooling water system other than the water system (for example, a system that supplies cooling water to the heat exchanger of the reactor purification system). The arithmetic unit 37A creates image information including the measurement values of the dew point meter 18 and the drain water meter 38 together with the determination information (message information “leakage occurred in the auxiliary machine cooling system”), and this image information is displayed on the display device 37B. Output to. This image information is displayed on the display device 37B. The image information includes nuclide analysis information. The arithmetic unit 37A warns the operator when outputting the message information “leak generation in reactor cooling system”, “leak generation in steam system” or “leak generation in auxiliary cooling system” to display device 37B. Sound a buzzer (not shown).

  FIG. 9 shows an example of information displayed on the display device 37 </ b> B when a leak occurs in the steam system in the reactor containment vessel 1. Dew point temperature measurement value, radioactivity measurement value, nuclide analysis value of radioactive nitrogen component (N-13, N-16), nuclide analysis value of corrosive biological component (Mn-54, Co-60), measurement value of drain water amount And a measured value of hydrogen concentration (not shown) are displayed. The display information shown in FIG. 9 is created by the arithmetic device 37A. Since the nuclide analysis values of N-13 and N-16 rise and the nuclide analysis values of Mn-54 and Co-60 hardly change, this display information indicates that leakage occurs in the vapor system. The display device also displays message information of “leak occurrence in steam system”. The operator can know that leakage has occurred in the steam system by looking at the information displayed on the display device.

  Generally, the time required for the steam to reach from the RPV 4 to the condenser is about 1 minute. Although depending on the value of the N-13 / N-16 ratio in the RPV 4, after the occurrence of the leakage, the nuclide analysis and analysis by the data processing device 36 can be promptly performed to display the message information of the leakage. By storing each piece of information obtained by the wave height analyzer 34A in the storage device 35, the processing by the data processing device 36 can be performed smoothly.

  Although leakage monitoring in the reactor containment vessel 1 has been described, leakage monitoring in the turbine building 2 using the LDS 20 will be described. The control device 36c sequentially performs the switching operation of the selector valve 14B. The gas sampled sequentially from the sampling ports 21a, 21b, 21c, 21d, 21e,... Through the corresponding sampling pipe 52 by the drive of the blower 29 is sent to the measurement chamber 24B through the pipe line 28, and the Ge detector 16B. Detects γ rays. The γ-ray detection signal output from the Ge detector 16B is amplified by the amplifier 33B and input to the wave height analyzer 34B. The wave height analyzer 34B performs the same processing as the wave height analyzer 34A, and outputs the obtained information to the data processing device 36. The data processing device 36 executes the processing shown in FIG. 8 as described above, and leakage of the steam system (for example, the main steam pipe 10 and the instrumentation pipe connected to the main steam pipe 10) in the turbine building 2. Message information of the reactor or the message information of leakage of the reactor cooling water system (for example, the water supply pipe 11 and the instrumentation pipe connected to the water supply pipe 11). These message information is displayed on the display device 37b. The feed water flowing through the feed water pipe 11 contains trace amounts of Mn-54 and Co-60. These are contained in the steam discharged to the main steam pipe 11, and the condensed water of the steam extracted by the feed water heater (not shown) from the turbine by the condensation of the steam in the condenser 8. By introducing it into the condenser 8, it enters the water supply. For this reason, the data processor 36 can perform the same process as the leakage in the reactor containment vessel 1. The estimation of the position of the leak location in the steam system using the N-13 / N-16 ratio is similarly performed.

  This example shows the amount of radioactivity of radioactive nitrogen (for example, N-13, N-16) or radioactive carbon (for example, C-15) produced by the nuclear reaction obtained by nuclide analysis of the γ-ray detection signal and Since the amount of radioactivity of corrosion products (for example, Mn-54, Co-58, Co-60) generated in the cooling water in the RPV 4 is monitored, the inside of the reactor containment vessel 1 (or the turbine building 2) It can be accurately determined in a short time whether leakage has occurred in either the steam system or the reactor coolant system. For this reason, the operator can quickly know where the leakage has occurred, and can prepare for maintenance monitoring and repair work of the nuclear power plant without delay. The safety of nuclear power plants is significantly improved.

  In this example, when leakage occurs in the vapor system, the ratio of the radioactivity amounts of two radionuclides having different half-lives obtained based on information obtained by nuclide analysis, for example, N-13 / N- Since the leak location is estimated using the 16 ratio (or N-16 / C-15 ratio or N-13 / C-15 ratio), the position of the leak location in the steam system can be specified with higher accuracy. . For this reason, the operator can quickly determine the propagation of the accident and can more appropriately prepare for the repair work. Since the storage device 35 stores information that associates the N-13 / N-16 ratio (or N-16 / C-15 ratio or N-13 / C-15 ratio) with the leakage location, N-13 It is possible to easily identify the leaking point in the steam system using the / N-16 ratio or the like.

  If it is determined that “a leak has occurred in the reactor cooling system”, it is connected to any system of the reactor cooling system, that is, the recirculation system, the reactor purification system, the residual heat removal system, the water supply pipe, and one of these. The operator needs to know which system of each of the instrumentation pipes in which the leakage has occurred. In this embodiment, by switching the selector valve 14A (or selector valve 14B), the recirculation system, the reactor purification system, the residual heat removal system, the feed water pipe, and the pipes for instrumentation that are branch pipes of these are close to each other. The sampling ports are respectively arranged, and the gas sampled from each sampling port is sequentially supplied to the measurement chamber 24A (or measurement chamber 24B). For this reason, based on the amount of radioactivity detected by the Ge detector 16A (or Ge detector 16B), a system (for example, instrumentation piping) in which leakage occurs in the reactor cooling water system can be specified. That is, the system near the sampling port that samples the gas in which the highest amount of radioactivity is detected leaks. The computing device 37A creates information on the system name of the system in which the leakage of the reactor cooling water system has occurred, and outputs it to the display device 37B for display.

  The leakage of the reactor cooling water system can also be detected by measuring the Co-58 γ rays with a Ge detector. In addition, the location of leakage in the reactor coolant system is identified by creating database information similar to that shown in FIG. 7 by using two corrosion products having different half-lives (for example, Mn-54 and Co-60). By doing so, it can be performed in the same manner as the steam system.

  The LDS of the nuclear facility according to embodiment 2 which is another embodiment of the present invention will be described with reference to FIG. The LDS 20A of the present embodiment has a configuration in which the radioactivity measurement units 30A and 30B of the LDS 20 of the first embodiment are replaced with radioactivity measurement units 30C, respectively, and the other configurations are the same as the LDS 20. The radioactivity measurement unit 30C for the reactor containment vessel 1 is provided in the pipeline 23, and the radioactivity measurement unit 30C for the turbine building 2 is provided in the pipeline 28. The radioactivity measurement unit 30C includes a Ge detector 16A, an amplifier 33A, and a measurement chamber 24C. The measurement chamber 24C is installed in the pipelines 23 and 28, respectively. The Ge detector 16A connected to the amplifier 33A is installed near the measurement chamber 24C. The amplifier 33A of the radioactivity measuring unit 30C for the reactor containment vessel 1 is connected to the wave height analyzer 34A, and the amplifier 33A of the radioactivity measuring unit 30C for the turbine building 2 is connected to the wave height analyzer 34B.

  The LDS 20 </ b> A also performs leakage monitoring in the same manner as the LDS 20, and can obtain the effects produced by the LDS 20.

  The LDS of the nuclear facility according to embodiment 3 which is another embodiment of the present invention will be described with reference to FIG. Before that, an example of a nuclide analysis measurement spectrum will be described with reference to FIG. An example of a nuclide analysis measurement spectrum as a result of nuclide analysis of the gamma ray detection signal output from the Ge detector 16A capable of covering the low energy region to the high energy region of γ rays is shown in FIG. 3 as described above. . The Ge detector 16A can detect γ rays in a wide energy range from 0 MeV to 8 MeV. However, when detecting γ-rays in a low energy region in a high energy measurement range, the measurement region is reduced, so that the accuracy of nuclide analysis is relatively lowered. FIG. 11A shows a nuclide analysis measurement spectrum in the low energy region, and FIG. 11B shows a nuclide analysis measurement spectrum in the high energy region. Thus, by detecting separately the γ rays in the high energy region and the γ rays in the low energy region, a more accurate nuclide analysis measurement spectrum can be obtained in each region.

  The LDS 20B of the present embodiment shown in FIG. 12 has a configuration in which the radioactivity measurement units 30A and 30B of the LDS 20 of the first embodiment are replaced with radioactivity measurement units 30D, respectively, and the other configurations are the same as the LDS 20. The radioactivity measurement unit 30D is provided with two Ge detectors 16C and 16D in the measurement chamber 24D. The Ge detector 16C is a radiation detector for a low energy region (0 to 2 MeV), and the Ge detector 16D is a radiation detector for a high energy region (4 to 8 MeV). The Ge detector 16D for the high energy region has a large sensitive volume, and the Ge detector 16C for the low energy region has a small sensitive volume. Similarly to the radioactivity measurement unit 30A, the radioactivity measurement unit 30D includes a roll filter 15A in the measurement chamber 24D. In the radioactivity measurement unit 30D for the reactor containment vessel 1, the Ge detector 16C is connected to the pulse height analyzer 34A via the amplifier 33C, and the Ge detector 16D is connected to the pulse height analyzer 34A via the amplifier 33D. In the radioactivity measurement unit 30D for the turbine building 2, the Ge detector 16C is connected to the wave height analyzer 34B via the amplifier 33C, and the Ge detector 16D is connected to the wave height analyzer 34B via the amplifier 33D.

  Since the present embodiment includes the radioactivity measurement unit 30D having the Ge detectors 16C and 16D, the γ rays in the low energy region and the high energy region are detected by the respective Ge detectors adapted to those regions. The accuracy of the nuclide analysis results for the low energy region and the high energy region is improved as compared with the analysis results in the first embodiment. For this reason, leakage in the steam system and the reactor coolant system can be detected with higher accuracy. Further, in this embodiment, it is possible to specify the leakage location in the steam system using the N-13 / N-16 ratio or the like with higher accuracy than in the first embodiment.

  An LDS of a nuclear facility according to embodiment 4 which is another embodiment of the present invention will be described with reference to FIG. The LDS 20C of the present embodiment has a configuration in which the Ge detector 16A of the second embodiment is replaced with the Ge detectors 16C and 16D described in the third embodiment, and other configurations are the same as the LDS 20A of the second embodiment. Since this embodiment also includes the Ge detectors 16C and 16D, the effect produced in the third embodiment can be obtained.

An LDS of a nuclear facility according to embodiment 5 which is another embodiment of the present invention will be described with reference to FIG.
The LDS 20D of the present embodiment has a configuration in which the radioactivity measurement units 30A and 30B of the LDS 20 of the first embodiment are replaced with radioactivity measurement units 30E, and the wave height analyzers 34A and 34B are replaced with multi-wave height analyzers 45, respectively. Other configurations are the same as those of the LDS 20. In FIG. 14, the measurement chamber is omitted.

  The radioactivity measurement unit 30E includes a Ge detector 16A, a preamplifier 42, and amplifiers 43 and 44. The amplifiers 43 and 44 are connected to the preamplifier 42 connected to the Ge detector 16 </ b> A and further connected to the multi-wave height analyzer 45. The amplifiers 43 and 44 are amplifiers having gains suitable for the low energy region and the high energy region, respectively. The γ-ray detection signal output from the Ge detector 16A is amplified by the preamplifier 42 and then input to the amplifiers 43 and 44, respectively. The respective γ-ray detection signals amplified and output by the amplifiers 43 and 44 are input to the multi-wave height analyzer 45 and simultaneously subjected to nuclide analysis.

  Also in this embodiment, the effect produced by the LDS 20B of the third embodiment provided with the Ge detectors 16C and 16D can be obtained. Furthermore, since only one Ge detector is required as compared with the third embodiment, the configuration of the LDS can be simplified as compared with the LDS 20B.

An LDS of a nuclear facility according to embodiment 6 which is another embodiment of the present invention will be described with reference to FIG.
The LDS 20E of the present embodiment has a configuration in which the radioactivity measurement unit 30E of the LDS 20D of the fifth embodiment is replaced with a radioactivity measurement unit 30F, and other configurations are the same as the LDS 20D. In FIG. 15, the measurement chamber is omitted.

  The radioactivity measurement unit 30F includes a Ge detector 16A, a preamplifier 42, a bias amplifier 46, and an amplifier 48. The preamplifier 42 is connected to the Ge detector 16A. Each of the bias amplifier 46 and the amplifier 48 is connected to the preamplifier 42 and further connected to the multi-wave height analyzer 45. The amplifier 48 is a gain amplifier for a low energy region (0 to 2 MeV). The bias amplifier 46 is an amplifier having a gain suitable for a high energy region (4 to 8 MeV). The γ-ray detection signal output from the Ge detector 16A is amplified by the preamplifier 42 and input to the bias amplifier 46 and the amplifier 48, respectively. Each γ-ray detection signal amplified and output by the bias amplifier 46 and the amplifier 48 is input to the multi-wave height analyzer 45 and simultaneously subjected to nuclide analysis.

  Also in this embodiment, the effect produced by the LDS 20B of the third embodiment provided with the Ge detectors 16C and 16D can be obtained. Furthermore, since only one Ge detector is required as compared with the third embodiment, the configuration of the LDS can be simplified as compared with the LDS 20B. In this embodiment, nuclide analysis can be performed in the low energy region and the high energy region with the same energy resolution.

  The LDSs of the above-described embodiments can be applied not only to boiling water nuclear power plants but also to other nuclear facilities such as pressurized water nuclear power plants, reprocessing facilities, and high-level waste handling facilities. .

BRIEF DESCRIPTION OF THE DRAWINGS It is a block diagram of the leakage monitoring system of the nuclear facility of Example 1 which is one preferable Example of this invention. It is a detailed block diagram of each radioactivity measurement part shown in FIG. It is explanatory drawing which shows an example of the result of the nuclide analysis performed with the wave height analyzer shown in FIG. It is explanatory drawing which shows each attenuation | damping curve of N-13 and N-16. It is explanatory drawing which shows attenuation with the elapsed time of N-13 / N-16 ratio. It is a typical block diagram which shows the connection state of the branch pipe in main steam piping, and instrument piping. It is explanatory drawing which shows an example of the database information which shows the relationship between each component installation place and N-13 / N-16 ratio. 2 is a flowchart of processing procedures executed by the data processing apparatus shown in FIG. It is explanatory drawing which shows an example of the image information displayed on the display apparatus shown in FIG. It is a block diagram of the radioactivity measurement part in the leakage monitoring system of the nuclear facility of Example 2 which is another Example of this invention. An example of nuclide analysis is shown, (A) is explanatory drawing which shows the example of nuclide analysis in a low energy area | region, (B) is explanatory drawing which shows the example of nuclide analysis in a high energy area | region. It is a block diagram of the radioactivity measurement part in the leakage monitoring system of the nuclear facility of Example 3 which is another Example of this invention. It is a block diagram of the radioactivity measurement part in the leakage monitoring system of the nuclear power plant of Example 4 which is another Example of this invention. It is a block diagram of the radioactivity measurement part in the leakage monitoring system of the nuclear facility of Example 5 which is another Example of this invention. It is a block diagram of the radioactivity measurement part in the leakage monitoring system of the nuclear facility of Example 6 which is another Example of this invention.

Explanation of symbols

  DESCRIPTION OF SYMBOLS 1 ... Reactor containment vessel, 2 ... Turbine building, 3 ... Core, 4 ... Reactor pressure vessel, 5 ... Recirculation piping, 7 ... Turbine, 8 ... Condenser, 9 ... Feed pump, 10 ... Main steam pipe, 11 ... water supply pipe, 13, 13a, 13b, 13c, 13d, 13e, 21a, 21b, 21c, 21d, 21e ... sampling port, 14A, 14B ... selector valve, 15A, 15B ... roll filter -, 16A, 16B, 16C, 16D ... Ge detector, 18 ... dew point meter, 20, 20A, 20B, 20C, 20D, 20E ... leak monitoring system, 24A, 24B, 24C, 24D ... measurement chamber, 25, 26 ... Roll, 27 ... Strip filter, 30A, 30B, 30C, 30D, 30E, 30F ... Radioactivity measurement unit, 34A, 34B ... Wave height analyzer, 35 ... Storage device, 36 ... Data processing device 37 ... Operation panel, 37A ... Arithmetic unit, 37B ... Display device, 37C ... Control device, 38 ... Drain water meter, 42 ... Preamplifier, 43, 44, 48 ... Amplifier, 45 ... Multi-wave height analyzer, 46 ... Bias amplifier, 49, 52 ... Sampling piping.

Claims (17)

  A semiconductor radiation detector for detecting radiation from a radionuclide contained in a gas guided by a plurality of sampling pipes each having a sampling port disposed at a plurality of locations in a nuclear facility having a steam system pipe and a reactor cooling water system pipe; , A nuclide analysis device that performs nuclide analysis using a radiation detection signal output from the semiconductor radiation detector, and nuclide analysis information from the nuclide analysis device are input, and radioactive nitrogen contained in the nuclide analysis information and A leakage location specifying device for determining whether a leakage location is in the steam system piping and the reactor cooling water system piping based on at least one of the radioactive carbon activity and the corrosion product activity Leakage monitoring system for nuclear facilities characterized by this.   Radiation from radionuclides contained in a plurality of sampling pipes each having sampling ports arranged at a plurality of locations in a nuclear facility including a steam system pipe and a reactor cooling water system pipe, and a gas guided by each of the sampling pipes A semiconductor radiation detector to be detected, a nuclide analyzer that performs nuclide analysis using a radiation detection signal output from the semiconductor radiation detector, and nuclide analysis information from the nuclide analyzer are input, and this nuclide analysis information Leakage that determines whether the leak location is in the steam system piping and the reactor cooling water system piping based on the radioactive activity of at least one of radioactive nitrogen and radioactive carbon and the radioactive activity of corrosion products A leakage monitoring system for a nuclear facility, characterized by comprising a location identifying device.   The nuclear facility leakage monitoring system according to claim 2, wherein the plurality of sampling pipes are disposed in a nuclear reactor containment vessel of the nuclear power facility having the steam pipe and the reactor cooling water pipe inside.   The nuclear facility leakage monitoring system according to claim 2, wherein the plurality of sampling pipes are disposed in a turbine building of the nuclear power facility having the steam pipe and the reactor cooling water pipe inside.   The leak location identifying device includes a radioactivity amount of a first radionuclide selected from N-13, N-16, and C-15 having different half-lives included in the nuclide analysis information, and among them. The leakage monitoring of a nuclear facility according to any one of claims 1 to 4, wherein a position of the leakage portion in the steam system piping is specified based on a ratio of the radioactivity amounts of the selected second radionuclide. system.   The ratio of the radioactivity of the first radionuclide and the radioactivity of the second radionuclide is the ratio of the radioactivity of N-13 to the radioactivity of N-16, and the ratio of the radioactivity of C-15 to the radioactivity of N-16. 6. The leakage monitoring system for a nuclear facility according to claim 5, wherein the leakage monitoring system is one selected from the ratio of the C-15 activity to the N-13 activity.   As the semiconductor radiation detector, a first semiconductor radiation detector that detects radiation in a first energy region, and a second semiconductor radiation detector that detects radiation in a second energy region higher in energy than the first energy region. The leakage monitoring system for a nuclear facility according to any one of claims 1 to 6, wherein the leakage monitoring system is used. A radiation detection signal output from the semiconductor radiation detector is input, a first amplification device for amplifying the radiation detection signal in a first energy region, and a radiation detection signal output from the semiconductor radiation detector are input, A second amplifying device for amplifying the radiation detection signal in a second energy region having a higher energy than the first energy region;
7. The leakage monitoring of a nuclear facility according to claim 1, wherein the nuclide analyzer performs nuclide analysis by inputting respective output signals of the first amplifying device and the second amplifying device. 8. system.   9. The apparatus according to claim 1, further comprising: a measurement chamber provided with the semiconductor radiation detector to which the gas is introduced; and a selector valve that switches connection between the plurality of sampling pipes and the measurement chamber. A leakage monitoring system for nuclear facilities as described in the paragraph.   The nuclear facility leakage monitoring system according to any one of claims 1 to 9, wherein a filter device that captures the radionuclide is provided in the measurement chamber.   The leakage monitoring system for a nuclear facility according to any one of claims 1 to 10, further comprising a display device that displays position information of the leakage location specified by the leakage location specifying device.   An image information creation device that creates image information including hydrogen concentration meter, dew point meter and drain water meter installed in the nuclear facility, each measurement information measured by these measurement devices, and the nuclide analysis information, The leakage monitoring system for a nuclear facility according to any one of claims 1 to 10, further comprising a display device that displays the image information.   Radiation from the radionuclide contained in the gas sampled from the nuclear facility equipped with the steam system piping and the reactor cooling water system piping is detected, and the nuclide analysis is performed using the radiation detection signal obtained by this detection. Based on the radioactivity of at least one of radioactive nitrogen and radiocarbon and the radioactivity of the corrosion products included in the nuclide analysis information obtained in the analysis, whether the leakage point is in the steam system piping and the reactor cooling A method for monitoring leakage of a nuclear facility, characterized in that it is determined whether the pipe is in a water system.   The nuclear facility leakage monitoring method according to claim 13, wherein the sampled gas is a gas sampled in a nuclear reactor containment vessel of the nuclear facility.   The nuclear facility leakage monitoring method according to claim 13, wherein the sampled gas is a sampled gas in a turbine building of the nuclear facility.   Radioactivity amount of the first radionuclide selected from N-13, N-16 and C-15 having different half-lives and the second radionuclide selected from them, which are included in the nuclide analysis information The nuclear facility leakage monitoring method according to any one of claims 13 to 15, wherein a position of the leakage portion occurring in the steam system piping is specified based on a ratio of the amount of radioactivity.   The nuclear facility leakage monitoring method according to any one of claims 13 to 16, wherein the radiation is detected by a semiconductor radiation detector.

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