JP2011137700A - Leakage detector - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To determine the presence or absence of a leakage in a containment vessel during nuclear reactor operation with high accuracy. <P>SOLUTION: The leakage detector 26 detects a leakage of main steam or reactor water of a nuclear reactor by detecting and analyzing radiation from radionuclide contained in the main steam and the reactor water. By using a radiation detection signal output from a gamma ray detector 4, a relative gamma ray counting rate of sampling gas which passed through a gas chemical form separator 7 to that which did not pass through the gas chemical form separator 7 is calculated. Determination as to whether there is a leakage or not and whether the leakage is main steam leakage or reactor water leakage can be made with high accuracy depending on which of the preliminarily memorized main steam leakage pattern and reactor water leakage pattern is closer to the calculated gamma ray counting rate. <P>COPYRIGHT: (C)2011,JPO&INPIT

Description

  The present invention relates to a technique for detecting leakage in a nuclear reactor containment vessel (hereinafter also simply referred to as a “containment vessel”), and in particular, determination of the presence or absence of leakage and whether or not the leakage occurred is a main steam or reactor. The present invention relates to a leak detection technique suitable for determining whether water is present or specifying a leak position in a containment vessel.

  Conventionally, in a leak detection device in a containment vessel of a nuclear power plant, for example, for a gas sampled from one place in a containment vessel during operation of a nuclear reactor, a gross gamma (γ Measure and calculate the integrated energy of the line).

  Moreover, in patent document 1, the radionuclide (N-13, N-16, Mn-54, gas in the gas sampled from the some sampling port arrange | positioned in the containment vessel near the steam system and the reactor cooling water system. Co-60) γ-rays are detected by a Ge detector, and a wave height analyzer performs nuclide analysis based on the γ-ray detection signal. Based on the radioactivity amount of N-13, Mn-54, etc. in the nuclide analysis information. A technique is disclosed in which a data processing device determines whether a leak is a steam system or a reactor cooling water system.

JP 2008-96345 A

  However, in the above-described prior art, even if it is possible to confirm that a leak has occurred, the dew point in the containment vessel can be used to confirm that this is a real leak rather than a malfunction of the device. It was necessary to make a comprehensive judgment based on various information such as the amount of condensed water from the containment vessel. In addition, since the leak location cannot be identified, there is a problem that it takes time to perform an on-site investigation in the containment vessel after shutting down the reactor to identify the leak location.

  Therefore, the present invention has been made in view of the above-described problems, and it is an object of the present invention to accurately determine whether or not there is a leak in the containment vessel during operation of the nuclear reactor and to specify the leak position.

  In order to solve the above-mentioned problems, the present invention provides a leak detection device for detecting leakage of main steam or reactor water by detecting and analyzing radiation from radionuclides contained in reactor main steam and reactor water. is there. Using the radiation detection signal output from the gamma ray detector, the relative gamma ray count rate of the sampling gas when passing through the gas chemical form separator is calculated with respect to the case where the gas chemical form separator is not passed through. Whether the main steam leak pattern or the reactor water leak pattern is close to the stored main steam leak pattern or whether it is a main steam leak or a reactor water leak can be determined with high accuracy.

  According to the present invention, it is possible to accurately determine the presence or absence of leakage in the containment vessel during the operation of the nuclear reactor, and to specify the leakage position.

It is a block diagram of the leak detection apparatus of 1st Embodiment of this invention. It is a block diagram of the leak detection apparatus of 2nd Embodiment of this invention. It is a block diagram of the leak detection apparatus of 3rd Embodiment of this invention. It is the example of installation of the leak detection apparatus of 1st Embodiment of this invention. It is the other example of installation of the leak detection apparatus of 1st Embodiment of this invention. It is a figure which shows an example of the time change of the count rate in the leak detection apparatus at the time of the reactor water leak in 1st Embodiment of this invention. It is a figure which shows another example of the time change of the count rate in the leak detection apparatus at the time of the main steam leak in 1st Embodiment of this invention. It is a figure which shows another example of the time change of the count rate in the leak detection apparatus at the time of leaking in the vicinity of the sampling port a in 1st Embodiment of this invention. It is a figure which shows the other example of the time change of the count rate in the leak detection apparatus at the time of leaking in the intermediate part of the sampling port a and the sampling port b in 1st Embodiment of this invention.

  Hereinafter, modes for carrying out the present invention (hereinafter referred to as embodiments) will be described with reference to the drawings (refer to drawings other than the referenced drawings as appropriate). In this embodiment, nitrogen 13 is taken as an example as an index nuclide for leakage. Nitrogen 13 is present in a large amount in the main steam and reactor water in the containment vessel, and is not present otherwise, so it can be determined that leakage has occurred by detecting nitrogen 13. Further, the nitrogen 13 exists mainly in the chemical forms of nitric oxide and ammonia in the main steam, and exists mainly in the chemical forms of nitric acid, nitrogen dioxide and ammonia in the reactor water. Therefore, the concentration of nitrogen 13 in the sampling gas (hereinafter, also simply referred to as “gas”) is measured mainly by direct measurement and measurement after passing through the gas chemical form separator. It is possible to specify whether steam leaks or reactor water leaks (details are given later). Nitrogen 13 is more suitable as an indicator nuclide because of its long half-life of about 10 minutes, but oxygen 19, nitrogen 16, fluorine 18, etc. may also be used as an indicator nuclide.

  Further, a dehumidifying cooler and a ventilator are installed in the containment vessel in order to keep the temperature and humidity in the containment vessel substantially uniform, and there is a gas flow. The radionuclide in the leakage component moves according to the flow, but the concentration is high in the vicinity of the leakage, and the concentration is considered to decrease as the distance increases. Therefore, it is possible to specify the approximate leak position by comparing the concentrations of radionuclides in the gas sampled from a plurality of positions.

(First embodiment)
As shown in FIG. 1, a leak detection device 26a (26) according to the first embodiment includes a gas cell 1 that contains a sampling gas 10 (hereinafter, also referred to as “sampling gas” without reference numerals), and from outside the gas. Shield 2 for shielding background radiation, piping 3 for introducing sampling gas into gas cell 1, gamma ray detector 4 (radiation detector) for measuring the concentration of radionuclide in gas cell 1, CPU (Central Processing Unit) A data collection personal computer 5 that collects data from the gamma ray detector 4, a high voltage power source 6 for applying a high voltage to the gamma ray detector 4, and a sampling gas Gas chemical form separator 7 for removing only radionuclide having a specific chemical form, and sampling gas is passed through gas chemical form separator 7 And a solenoid valve 8 for switching between line-impermeable and in.

  The gamma ray detector 4 discriminates the energy from the radionuclide in the sampling gas, and measures the radionuclide concentration using the radionuclide normally present only in the main steam or the reactor water as an indicator nuclide (for example, nitrogen 13). Further, the sampling exhaust gas 11 is discharged from the gas cell 1.

  Referring to FIG. 6, an example of the gamma ray count rate (radiation count rate) of the indicator nuclide in the leak detector 26 in the gas chemical form separator unpassed gas and in the gas chemical form separator pass gas when the reactor water leaks explain. Since the nitrogen 13 in the reactor water exists in the chemical forms of nitric acid, nitrogen dioxide and ammonia, when the reactor water leaks, the containment vessel is in accordance with the saturated vapor pressure determined by the chemical forms of nitric acid, nitrogen dioxide and ammonia. It becomes the gas component inside.

  When a water bubbling device is used as the gas chemical form separator 7, each chemical substance of nitric acid, nitrogen dioxide, and ammonia is easily dissolved in water, so that it is trapped by the gas chemical form separator 7, and the gas chemical form separator is used. In the gas after passing through 7, almost no nitrogen 13 remains. Therefore, by measuring and comparing the gamma ray count rate of nitrogen 13 in the gas that has passed through the gas chemical form separator 7 and the gas that has not passed through, it can be identified as reactor water leakage.

  With reference to FIG. 7, an example of the gamma ray count rate of the indicator nuclide in the leakage detector 26 in the gas chemical form separator non-passing gas and the gas chemical form separator passing gas at the time of main steam leakage will be described. Since the nitrogen 13 in the main steam exists in the form of ammonia and nitric oxide, when the main steam leaks, it becomes a gas component in the containment vessel in the chemical form of ammonia and nitric oxide.

  When a water bubbling device is used as the gas chemical form separator 7, ammonia is easily dissolved in water, but since nitric oxide hardly dissolves in water, only the form of ammonia is trapped in the gas chemical form separator 7, Nitrogen 13 in the form of nitric oxide also remains in the gas after passing through the gas chemical form separator 7. Therefore, by measuring and comparing the gamma ray count rate of nitrogen 13 in the gas that has passed through the gas chemical form separator and in the gas that has not passed through, it is confirmed that the gamma ray of nitrogen 13 is being measured even after passing through. Can be identified as leaking.

  That is, by using the radiation detection signal output from the gamma ray detector 4, the relative gamma ray count rate of the sampling gas when the gas chemical form separator is not passed through the gas chemical form separator 7 is obtained. The main steam is calculated and determined by comparing with the relative gamma ray count rate of the main steam and the relative gamma ray count rate of the reactor water stored in the storage means of the data collection personal computer 5. It is possible to determine whether the leakage of the furnace or the leakage of the reactor water.

  Next, with reference to FIG. 4, the installation example of the leak detection apparatus 26 of 1st Embodiment is demonstrated. Inside the storage container 30, a pressure container 31 and a gamma ray shield 32 are installed. Sampling gas in the storage container 30 from the sampling ports 35 installed at a plurality of positions (for example, about five places) in the storage container 30, and which sampling by the sampling gas switch 25 installed inside or outside the storage container 30 The gas sampled from the mouth 35 is specified and sent to the leak detector 26, and the gamma ray count rate from the radionuclide in the gas is measured. The sampling exhaust gas 11 discharged from the leak detection device 26 is returned to the inside of the storage container 30 from the sampling exhaust gas inlet 36.

  Referring to FIG. 8, gas sampled from sampling port a and other sampling ports b (not shown) when leakage occurs in the vicinity of sampling port a (not shown) which is one of sampling ports 35. An example of the gamma ray count rate in the leakage detection device 26 from the index nuclides inside will be described. The gamma ray count rate in the leak detection device 26 of the index nuclide in the gas sampled from the sampling port a in the vicinity of the leak is higher than the gamma ray count rate in the leak detection device 26 of the index nuclide in the gas sampled from the sampling port b. It is a value.

  Next, referring to FIG. 9, in the gas sampled from the sampling port a and the sampling port b when a leak occurs at an intermediate position between the sampling port a which is one of the sampling ports 35 and the other sampling port b. An example of the measurement of the gamma ray count rate in the leak detection device 26 from the index nuclide will be described. The gamma ray count rate in the leak detection device 26 of the index nuclide in the gas sampled from each of the sampling port a and the sampling port b is substantially equal.

  From these facts, the approximate leak position can be specified by comparing the gamma ray count rate in the leak detection device 26 of the index nuclide in the gas sampled from the plurality of sampling ports 35. That is, the gamma ray count rate (or concentration) of the radionuclide in the sampling gas sampled from each sampling port 35 arranged at a plurality of positions is measured and compared, and the higher the value, the corresponding sampling port 35. It can be determined that leakage has occurred at a position close to. In addition, the airflow is usually swirling in the storage container 30, but the leakage position can be specified with higher accuracy by taking the airflow into account by three-dimensional fluid analysis.

(Second Embodiment)
As shown in FIG. 2, in the leak detection device 26b (26) of the second embodiment, the sampling gas 10 from the containment vessel 30 is branched into two sampling lines, and one of them is directly sent to the gamma-ray emission nuclide measuring device 20. The other is sent to another gamma-ray emission nuclide measuring device 20 through the gas chemical form separator 7 installed in the front stage of the gamma-ray emission nuclide measurement device 20.

  Since the gamma-ray emission nuclide measuring instrument 20 is the same as that shown in FIG. 1 and is not shown in FIG. 1, a gas cell 1 for accumulating the sampling gas 10, a shield 2 for shielding background radiation from the outside of the gas, A pipe 3 for introducing the sampling gas 10 into the gas cell 1, a gamma ray detector 4 for measuring the radionuclide concentration in the gas cell 1, a data collecting personal computer 5 for collecting data from the gamma ray detector 4, and a gamma ray detector 4 A high voltage power source 6 for applying a high voltage is provided. By using two gamma-ray emission nuclide measuring instruments 20, the gas that has passed through the gas chemical form separator 7 and the gas that has not passed through without switching the line by the electromagnetic valve 8 as in the first embodiment. Can be measured simultaneously.

(Third embodiment)
As shown in FIG. 3, in the leakage detection device 26 c (26) of the third embodiment, the sampling gas 10 from the storage container 30 is branched into two sampling lines, one is directly sent to the gas cell 1, and the other is gas cell 1. It is sent to another gas cell 1 through the gas chemical form separator 7 installed in the previous stage. A gamma ray detector 4 is installed at the center of the two gas cells 1, and a movable shield 2 a is installed between the gamma ray detector 4 and the gas cell 1. By moving each shield 2a alternately, the gamma ray count rate from the radionuclide in the two gas cells 1 can be measured by one gamma ray detector 4. Usually, the cost of the single gamma ray detector 4 is higher than that of the two shields 2a. Therefore, this configuration can reduce the cost.

(Fourth embodiment)
As shown in FIG. 5, in the installation example of the leakage detection device 26 of the fourth embodiment, a pressure vessel 31 and a gamma ray shield 32 are installed inside the storage container 30 (for the same configuration as FIG. 4). (Omitted description). The gas in the storage container 30 is sampled from sampling ports 35 installed at a plurality of positions in the storage container 30 and measured by a plurality of leak detection devices 26 installed outside the storage container 30. With this configuration, the index nuclides in the gas sampled from the plurality of sampling ports 35 can be measured simultaneously.

  As described above, according to the leakage detection device 26 of the first to fourth embodiments, gas is sampled from a plurality of positions in the containment vessel 30, radiation in the sampling gas is energy-discriminated, and the main steam or reactor water is collected. By detecting a radionuclide that is normally present only (for example, nitrogen 13) as an index nuclide, the presence or absence of leakage can be determined with high accuracy.

  Moreover, the approximate position of the leak can be specified by measuring and comparing the concentration of the radionuclide in the sampling gas sampled from each sampling position. Further, by distinguishing and measuring the chemical form of the radionuclide in the sampling gas, it is possible to identify whether the main steam leaks or the reactor water leaks. As a result, it is possible to greatly reduce the time for on-site inspection after the leakage is found and the reactor is shut down.

  In addition to the water bubbling device, the gas chemical form separator 7 may be realized by an adsorbing means for adsorbing a specific chemical form among the radionuclides contained in the sampling gas.

DESCRIPTION OF SYMBOLS 1 ... Gas cell, 2, 2a ... Shielding body, 3 ... Piping, 4 ... Gamma ray detector, 5 ... Data collection personal computer, 6 ... High voltage power supply, 7 ... Gas chemical form separator, 8 ... Solenoid valve, 10 ... Sampling gas , 11 ... Sampling exhaust gas, 20 ... Gamma-ray emission nuclide measuring device, 25 ... Sampling gas switch, 26 ... Leak detector, 30 ... Containment vessel, 31 ... Pressure vessel, 32 ... Gamma ray shield, 35 ... Sampling port, 36 ... Sampling exhaust gas inlet

Claims (7)

A leak detection device that detects leakage of main steam or reactor water by detecting and analyzing radiation from radionuclides contained in the reactor main steam and reactor water,
The same radionuclide is contained in the main steam and reactor water of the reactor, and the chemical form is different between the radionuclide in the main steam and the radionuclide in the reactor water,
A gas cell for sequentially storing each sampling gas sampled from each sampling port arranged at a plurality of positions in the containment vessel of the reactor;
A gas chemical form separator installed between the sampling port and the gas cell to remove a specific chemical form of the radionuclide contained in the sampling gas;
Radiation from the radionuclide contained in the sampling gas from which the radionuclide of the specific chemical form contained in the gas cell has been removed, and the gas cell without passing through the gas chemical form separator from the sampling port A radiation detector for respectively detecting radiation from the radionuclide contained in the sampling gas contained in
Relative radiation count rate of the main steam when not passing through the gas chemical form separator, and when not passing through the gas chemical form separator Storage means for storing a relative radiation count rate of the reactor water when passing through the gas chemical form separator;
Using the radiation detection signal output from the radiation detector, the relative radiation count rate of the sampling gas when passing through the gas chemical form separator is calculated with respect to the case not passing through the gas chemical form separator. The main steam leakage is determined by determining which is close to the relative radiation count rate of the main steam stored in the storage means and the relative radiation count rate of the reactor water. Determine whether the reactor water leaks, and measure and compare the concentration of the radionuclide in the sampling gas sampled from each of the sampling ports arranged at a plurality of positions, the higher the concentration, the more the corresponding Arithmetic means for determining that the leakage has occurred at a position close to the sampling port;
A leak detection device comprising: The piping that passes through the gas chemical form separator and the pipe that does not pass through the gas chemical form separator are provided between the sampling port and the gas cell. Leak detection device. Two gas cells are provided,
The gas chemical form separator is provided between the sampling port and one of the gas cells,
The gas chemical form separator is not provided between the sampling port and the other gas cell,
The radiation detector is provided between the two gas cells;
A movable shield is provided between the radiation detector and the two gas cells, and when the radiation detector detects the radiation emitted from the radionuclide in one of the gas cells, The leak detection device according to claim 1, wherein the shield on the corresponding side moves to a position where radiation emitted from the radionuclide does not reach the radiation detector. The leak detection device according to claim 1,
A leak detection apparatus comprising a plurality of sampling ports corresponding to the plurality of sampling ports, and simultaneously measuring a radiation count rate based on a radionuclide in a sampling gas from each of the sampling ports. The gas chemical form separator is a water bubbling device for removing a specific chemical form having high solubility in water among the radionuclides contained in the sampling gas. The leak detection device described. The leak detection apparatus according to claim 1, wherein the gas chemical form separator is an adsorbing unit that adsorbs a specific chemical form of the radionuclide contained in the sampling gas. The leak detection apparatus according to claim 1, wherein the radionuclide is any one of nitrogen 13, oxygen 19, nitrogen 16, and fluorine 18.

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