JP2005265856A - Method for detecting failed fuel assembly and its system - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a method for detecting failed fuel assembly and its system which can identify failed fuel assemblies by performing in-core shipping quickly and with high detection sensitivity. <P>SOLUTION: Irradiation days of the failed fuel assemblies are estimated and detection operation of fission product is conducted for fuel assemblies of burnup corresponding to the estimated irradiation days to identify the failed fuel assemblies. The detection operation is conducted by sampling a fluid sample having sufficiently accumulated radioactive nuclides discharged out of leakers with a shipper cap having a structure capable of tight contact fixing to fuel assembly outer wall 43 and attaching the top part 11 above a fuel grasping tool 30 of a fuel exchanger mast 20 and using, and measuring a radioactivity intensity of the sample. <P>COPYRIGHT: (C)2005,JPO&NCIPI

Description

  The present invention relates to a method and apparatus for detecting a damaged fuel assembly generated in a nuclear power plant, evaluating its burnup and identifying it.

  In a nuclear power plant, a fuel replacement method is adopted in which a portion (irradiated fuel) in which the burnup has reached a predetermined value among all nuclear fuel in the core is replaced with a new fuel every operation cycle. The burnup is evaluated for each operation cycle, and about 1/4 of the total fuel is replaced with new fuel. In that case, the nuclear fuel in the first cycle to the fourth cycle, that is, the nuclear fuel having different burnups in four stages is always loaded in the core except the initial operation.

  If the fuel cladding breaks down for some reason during the operation of the plant and the release of fission products (hereinafter referred to as FP) occurs, the relevant damaged fuel assembly will be replaced at any time or during such regular replacement. (Hereinafter referred to as “leaker”) is identified and burnup is evaluated and replaced with healthy fuel.

  In nuclear power plants, a broken fuel detection system is provided for the purpose of detecting the occurrence of such a leaker and identifying it, that is, knowing the position in the reactor. During normal operation, the radioactive concentration of gaseous waste (hereinafter referred to as off-gas) sampled by this damaged fuel detection system and the radioactive concentration of iodine (I) in the reactor water are constantly monitored, and these radioactive concentrations are monitored. The riser detects the occurrence of leaks. Usually, delayed neutrons, β rays, or γ rays emitted by FP are detected, and radioactivity concentration measurement, nuclide analysis, comparison with the background, and the like are performed.

  In general, regarding the detection of a leaker, a method that captures a change in radioactivity concentration of off-gas is considered to have the highest sensitivity. This is because, for example, a rare gas component such as krypton (Kr) or xenon (Xe) is most easily released into the reactor water from the pinhole of the fuel cladding as compared with other FPs, and is released from the fuel rod to the reactor water. This is because it has a chemical property that it immediately shifts to the vapor system and reaches the off-gas sampling system without reacting with other elements. Therefore, when the measured value of the off-gas radioactivity concentration increases, it can be estimated that the rare gas component of FP was released along with the occurrence of the leaker.

  On the other hand, I is an FP that is easily released from the fuel rod into the reactor water after the rare gas component, and has a chemical property that it is accumulated in the reactor water unlike the rare gas component. Therefore, when the measured value of the radioactivity concentration of the I isotope in the reactor water rises, it can be estimated that the release of I, which is FP, was caused by the occurrence of the leaker.

  Depending on the position of the generated leaker in the core, there are cases where a large amount of rare gas components are released and cases where a large amount of I is released. Therefore, in a nuclear power plant, the presence or absence of leakage is detected with higher sensitivity by capturing both the change in radioactivity concentration of off-gas and the change in radioactivity concentration of I in the reactor water.

  Conventionally, for example, in a light water reactor, a shipping method has been known as one of the so-called damaged fuel positioning methods for identifying which one of the loaded fuel assemblies is damaged in the plant where the occurrence of leakage is detected. ing. In this shipping method, the fuel assembly to be measured is isolated from the surroundings during the reactor shutdown period, and after a certain period of time, the fuel temperature rises due to decay heat, and then the I Measure the radioactivity by performing an operation called “shipping” to suck up the reactor water containing FP, or sample the FP of rare gases such as Kr and Xe diffused in the air while the reactor water is excluded by the air. It is a method to measure.

  In the measurement of these radionuclides in shipping, the method of measuring rare gas is highly sensitive because the background is small, as in the case of detection of the occurrence of leaker. In the measurement of I, the components present in the reactor water are considered as background. Therefore, sensitivity and accuracy are low.

  When isolating the fuel assembly to be measured from the surroundings and taking out the reactor water by shipping, usually, the upper part of the fuel assembly is covered with a cover called a sipper cap, and the air is fed into the sipper cap to supply the reactor in the fuel assembly. The radionuclide released from the broken fuel rod is accumulated in the reactor water by stopping the upward flow of water and isolating it from the surrounding area. Thereafter, the reactor water with the increased radionuclide concentration is drawn out, the radioactivity is measured, and the leaker is positioned. FIG. 9 shows a schematic cross section of an example of such a conventional sipper cap. In the figure, a conventional sipper cap 100 is used in a state of being gripped by a fuel gripper 200, is mounted so as to cover a fuel handle 300 at the top of a fuel assembly, and includes a sampling tube 400 and an air supply tube 500.

  In measuring I, there is a method that does not use a sipper cap. It is a method in which the reactor water is sampled and evaluated by making it easier to release I by pulling out the fuel body to reduce the water depth and lowering the pressure around the fuel body.

  Further, this shipping can be divided into in-core shipping performed in the reactor core without removing the fuel body and out-core shipping performed by removing the fuel body from the reactor core, depending on the location where the operation is performed. The former in-core shipping does not require the trouble of taking out the fuel body from the core, but the background of the reactor water becomes high, so the detection sensitivity and accuracy are inferior to the latter method. In addition, the latter out-core shipping is superior in detection sensitivity and accuracy as compared with the former, but requires a time margin for implementation because it requires a step of removing the fuel body from the core.

  Therefore, when the occurrence of leaker is detected, in-core shipping is first performed for identifying the leaker, and when identification cannot be performed by in-core shipping, the fuel body is moved to the adjacent fuel pool and further out-core shipping is performed. It is common to measure again.

  By the way, when the occurrence of a leaker is detected, it is conventionally performed to evaluate the burnup of the leaker prior to identification. This is because, as described above, the fuel loaded in the core is divided into several types according to the difference in the burnup, so when identifying the leaker by the above method, if the burnup of the leaker is known in advance, This is because the number of fuel assemblies to be shipped can be greatly reduced. For example, when it is configured to be replaced in the fourth cycle, it can be divided into four types according to the difference in burnup, that is, the number of fuel assemblies to be shipped can be reduced to ¼. it can. Thus, the burner evaluation of the leaker has a great influence on the subsequent plant operation.

  However, it is difficult to obtain information on the burnup of the leaker from routinely obtained measurements of radionuclides such as noble gases in off-gas and I isotopes in reactor water. The radionuclides that are always measured have a long half-life of several days, whereas the burnup evaluation requires the determination of nuclides with a half-life of several hundred days or more. This is because daily quantification of nuclides has not been made.

  Therefore, in a plant where the occurrence of leaker is detected, as a means for estimating the burnup of the target fuel assembly, conventionally, the radioactivity of Cs-137 and Cs-134 in the reactor water is measured by radiochemical analysis, and the radioactivity is measured. The burnup is evaluated from the change in the intensity ratio (Cs-134 / 137).

  However, the current leaky burnup evaluation method for evaluating burnup from this radioactivity intensity ratio (Cs-134 / 137) has the following drawbacks.

  First, Cs is less likely to be released from the fuel rod into the reactor water than its rare gas and I due to its chemical properties, and the radioactive concentration of Cs in ordinary reactor water is smaller than the lower detection limit. Therefore, a small-scale leaker has a very low radioactive concentration of Cs in the reactor water, and treatment of reactor water of 50 to 100 liters is required for the determination of Cs in the reactor water. Therefore, in the measurement, it is necessary not only to sufficiently remove interfering nuclides such as Co-60 and Na-24, but also to take measures to reduce the exposure due to these radionuclides.

  Moreover, the fact that the normal radioactivity concentration of Cs in the reactor water is smaller than the detection lower limit means that the error in the evaluation of the background included in the quantitative value of Cs obtained at the time of occurrence of the leaker is large in the burnup evaluation. It leads to things. Theoretically, if more reactor water is processed and analyzed, such errors can be reduced and the accuracy of the evaluation can be improved. However, if more than 100 liters of reactor water is to be analyzed, the interference nuclide Such a large amount of reactor water analysis is difficult to implement. In addition, in the case of a leaker in which the radioactive concentration of released Cs is smaller than the lower limit of detection, the burnup cannot be evaluated. In addition, since the Cs concentration in the reactor water further decreases after the plant is shut down, the current method for evaluating the burnup of leica using Cs has various difficulties, such as the fact that a reactor water sample must be collected during plant operation. It was.

  The conventional shipping method also has the following problems. First, since the sipper cap is used, the cap is attached and detached, the operation is performed to supply air into the sipper cap after installation, the reactor water flow rate is adjusted to isolate the reactor water in the fuel body, and the isolated reactor water. After the radionuclide is released from the damaged fuel body, a series of complicated operations of collecting and analyzing the reactor water is required.

  In particular, careful attention was required in the operation of supplying air into the sipper cap and adjusting the flow rate of the reactor water in order to isolate the reactor water. For example, if the cap is excessively pressurized during air supply, gas leaks from the gap between the cap and the fuel assembly, resulting in insufficient isolation of the reactor water, or the cap falls off. Inconveniences such as the need for re-mounting.

  Furthermore, in-core shipping, as described above, the measurement sample contains radionuclides in the reactor water, which is the background, so that the detection sensitivity is not so high and the adjustment of the reactor water flow rate is not limited to the above. The detection limit was also greatly affected. Specifically, when the reactor water flow rate is large, the reactor water is insufficiently isolated and the background increases, so the detection sensitivity is significantly deteriorated. It was enough.

  For this reason, the conventional shipping method using a sipper cap requires fine adjustment of the reactor water flow rate for sufficient isolation of the reactor water, which takes time and is therefore quick evaluation. It was hard to say that it was a suitable method for in-core shipping. Furthermore, a technique is required to quickly collect a measurement sample.

  On the other hand, in the method that promotes the release of radionuclides in the fuel body by changing the water depth around the fuel body by changing the water depth around the fuel body without using the sipper cap, the released radionuclide has a background effect. There was a problem that it became easier to receive and the detection limit value further deteriorated. In addition, as described above, there are cases where a large amount of I is released and a large amount of rare gas is released depending on the vertical position of the occurrence of the leaker, and the magnitude of damage does not necessarily correlate with the magnitude of the measured value. There was also a difficulty.

  The present invention has been made to cope with the above situation, and provides a method and apparatus for detecting a broken fuel assembly capable of identifying a broken fuel assembly by performing in-core shipping with high speed and high detection sensitivity. Is the purpose.

  The apparatus for detecting a damaged fuel assembly according to the first aspect of the present invention comprises a measuring means for detecting a fission product by measuring the radioactivity intensity of a sample, and is attached to the top of the target fuel assembly for detection operation. A sipper cap for isolating the fluid sample in the internal space, and a pipe for transferring the sample for connecting the sipper cap and the measurement means to transfer the isolated fluid sample to the measurement means. An apparatus for detecting a broken fuel assembly for identifying a body, wherein a top portion of the sipper cap is closely fixed to a predetermined position above a fuel gripper of a fuel changer mast, and an internal space of the sipper cap is the fuel The gripper has a shape capable of accommodating the fuel handle of the target fuel assembly held by the fuel gripper, and the sipper cap has a shape before being attached to the top of the target fuel assembly. The inner periphery of the lower edge of the sipper cap is closely fixed to the outer periphery of the side wall surface of the fuel channel of the target fuel assembly, and a contact fixing means is provided for storing and holding the isolated fluid sample in the internal space. It is said.

  According to a second aspect of the present invention, there is provided the apparatus for detecting a broken fuel assembly according to the first aspect, wherein the measurement means side of the sample transfer pipe is kept at a reduced pressure from the sipper cap side and stored in the internal space. It is further characterized by further comprising pressure adjusting means that enables the fluid sample to be transferred to the measuring means by a pressure difference between inside and outside.

  According to a third aspect of the present invention, the apparatus for detecting a damaged fuel assembly according to the present invention further comprises a gas introduction pipe for introducing an isolation gas into the inner space of the sipper cap according to the second aspect. The pipe for use has gas-liquid separation means for separating the liquid from the fluid sample to be transferred and taking out the gas sample.

  According to a fourth aspect of the present invention, there is provided the apparatus for detecting a damaged fuel assembly according to the third aspect, wherein the pipe for sample transfer removes the disturbing noble gas from the gas taken out by the gas-liquid separation means. It further has an interference noble gas removing means using as a gas sample.

  According to a fifth aspect of the present invention, there is provided the apparatus for detecting a damaged fuel assembly according to the fourth aspect of the present invention, wherein the fluid sample stored in the internal space and then transferred is a liquid sample.

  According to a sixth aspect of the present invention, there is provided the apparatus for detecting a damaged fuel assembly according to the fifth aspect, wherein the sample transfer pipe adsorbs and collects an I isotope in the liquid sample to be transferred. It is characterized by having I adsorbing means for separating.

  According to a seventh aspect of the present invention, there is provided a method for detecting a damaged fuel assembly in which a sipper cap is attached to the top of a target fuel assembly to be detected, and a fluid sample is isolated in the inner space thereof. A method of detecting a broken fuel assembly, wherein the fission product is detected and the broken fuel assembly is identified, wherein the top of the sipper cap is located above the fuel gripper of the refueling machine mast. The inner space of the sipper cap has a shape capable of accommodating the fuel gripper and a fuel handle of a target fuel assembly gripped by the fuel gripper; and the sipper The cap is fixed to the inner periphery of the lower edge of the sipper cap, which is attached to the top of the target fuel assembly, and is tightly fixed to the outer periphery of the side wall of the fuel channel of the target fuel assembly. A close contact fixing means for enabling storage in the internal space is provided, and when the fluid sample is isolated by attaching the sipper cap, the fuel handle of the target fuel assembly is gripped by the fuel gripper to target the sipper cap. It is attached to the top of the fuel assembly, and the inner periphery of the lower edge of the sipper cap is closely fixed to the outer periphery of the side wall of the fuel channel of the target fuel assembly, and the isolated fluid sample is stored in the inner space. Then, the fluid sample is transferred to the measuring means through the sample transfer pipe, and the radioactive intensity of the fluid sample is measured by the measuring means to detect the fission product, and the damaged fuel assembly is identified. It is said.

  According to an eighth aspect of the present invention, there is provided a method for detecting a damaged fuel assembly according to the seventh aspect of the present invention, wherein the sample transfer is performed when the fluid sample stored in the internal space is transferred to the measuring means via the sample transfer pipe. The measurement means side of the piping for piping is kept at a reduced pressure from the sipper cap side, and the fluid sample stored in the internal space is transferred to the measurement means by an internal and external pressure difference.

  A method for detecting a damaged fuel assembly according to a ninth aspect of the present invention is the method for detecting a damaged fuel assembly according to the eighth aspect of the present invention, wherein the fluid sample is stored in the internal space and then an isolation gas is introduced to separate the gas from the fluid sample to be transferred. Then, the gas sample is taken out, and then the total radioactivity intensity of the taken out gas sample is measured by the measuring means.

  A method for detecting a damaged fuel assembly according to a tenth aspect of the present invention is the method for detecting a damaged fuel assembly according to the eighth aspect, wherein after the fluid sample is stored in the internal space, an isolation gas is introduced and the gas is separated from the fluid sample to be transferred. The gas sample is taken out, and then the total β-ray intensity of the taken out gas sample is measured by the measuring means.

  A method for detecting a damaged fuel assembly according to an eleventh aspect of the present invention is the method for detecting a damaged fuel assembly according to the eighth aspect, wherein after the fluid sample is stored in the internal space, an isolation gas is introduced and the gas is separated from the fluid sample to be transferred. After that, the interference rare gas is removed, the gas sample is taken out, and the total radioactivity intensity of the gas sample is measured by the measuring means.

  A method for detecting a damaged fuel assembly according to a twelfth aspect of the present invention is the method for detecting a broken fuel assembly according to the eighth aspect of the present invention, wherein a fluid sample is stored in the internal space, an isolation gas is introduced, and the gas is separated from the transferred fluid sample. After that, the disturbing noble gas is removed, a gas sample is taken out, and the β-ray intensity of Kr-85 g in the gas sample is measured by the measuring means.

  According to a thirteenth aspect of the present invention, there is provided a method for detecting a damaged fuel assembly according to the eighth aspect, wherein after the fluid sample is stored in the internal space, an isolation gas is introduced and taken out from the fluid sample to be transferred. The radioactivity intensity of the Xe isotope in the gas is measured by the measuring means.

  According to a fourteenth aspect of the present invention, in the method for detecting a damaged fuel assembly according to the present invention, the total radioactivity intensity of the liquid sample stored in the internal space and then transferred is measured by the measuring means. It is characterized by.

  A method for detecting a damaged fuel assembly according to a fifteenth aspect of the present invention is the method for detecting a broken fuel assembly according to the eighth aspect of the present invention, wherein the nuclear type radioactivity intensity of the liquid sample stored in the internal space and then transferred is measured by the measuring means. It is characterized by that.

  A method for detecting a damaged fuel assembly according to a sixteenth aspect of the present invention is the method for detecting a broken fuel assembly according to the eighth aspect of the present invention, wherein the radioactivity intensity of the I isotope of the liquid sample stored in the internal space and then transferred is measured by the measuring means. It is characterized by measuring.

  According to a seventeenth aspect of the present invention, there is provided a method for detecting a damaged fuel assembly according to the present invention, according to the eighth aspect, wherein the I isotope is separated by adsorption from a liquid sample stored in the internal space and then transferred. The total radioactivity intensity of the I isotope is measured by the measuring means.

  The method for detecting a damaged fuel assembly of the present invention according to claim 18 is the method according to claim 8, wherein the I isotope is separated by adsorption from the liquid sample transferred after being stored in the internal space. It is characterized in that the radioactivity intensity of the I isotope is measured by the measuring means.

  According to a nineteenth aspect of the present invention, there is provided the method for detecting a damaged fuel assembly of the present invention according to the ninth, eleventh, fourteenth, or seventeenth aspect, wherein the measuring means is a GM type radiation detector.

  According to a twentieth aspect of the present invention, there is provided a method for detecting a damaged fuel assembly according to the present invention, wherein the measuring means is a coincidence type β-ray detector.

  According to a twenty-first aspect of the present invention, there is provided the method for detecting a damaged fuel assembly according to the thirteenth, fifteenth, sixteenth or eighteenth aspect, wherein the measuring means is a germanium semiconductor detector.

  A method for detecting a damaged fuel assembly according to a twenty-second aspect of the present invention is the method for detecting a broken fuel assembly according to the seventh and eighth aspects, wherein the radioactive intensity of the fission product in the gas sample stored in the internal space and then transferred and the liquid sample The radioactive intensity of the fission product therein is measured by the measuring means.

  The method for detecting a damaged fuel assembly according to a twenty-third aspect of the present invention is the method for detecting a damaged fuel assembly according to the twenty-second aspect, wherein the fission product of the gas sample is Kr-85 g and the fission product in the liquid sample is I-131. It is characterized by being.

  A method for detecting a damaged fuel assembly according to a twenty-fourth aspect of the present invention is the method according to the seventh or eighth aspect, wherein the fuel exchanger mast is moved up and down while the isolated fluid sample is stored and held in the internal space. The fuel assembly is extracted and inserted from a predetermined position, and then the fluid sample is transferred to the measuring means via the sample transfer pipe.

  As described above, according to the present invention, there is provided a method and apparatus for detecting a broken fuel assembly capable of identifying a broken fuel assembly by in-core shipping with high speed and high detection sensitivity. According to the present invention, since the same equipment piping can be used in common as one device, the burnup of the damaged fuel assembly can be more easily evaluated and identified.

  Furthermore, since many radionuclides can be continuously collected, separated, and measured as compared with the prior art, it is possible to cope well with various damage positions and scales.

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

  FIG. 1 shows the irradiation days of the fuel assemblies loaded in the reactor, and the Kr-85 g activity intensity ratio (Kr-85 g / m) to Kr-85 m in the radioactive gas waste generated during irradiation. It is a graph which shows the relationship. Kr-85m, which has a short half-life of 4.5 hours, becomes saturated as soon as combustion starts and takes a substantially constant value without significant fluctuations throughout the combustion period. On the other hand, since Kr-85g has a long half-life of 10 years, it accumulates in the fuel rod by combustion and increases its radioactivity intensity as combustion increases. The relationship shown in FIG. 1 exists between this radioactivity intensity ratio and the number of irradiation days, that is, the burnup. Therefore, if this radioactivity intensity ratio is known, the corresponding burnup is uniquely determined.

  By the way, Kr-85m and Kr-85g having different half-lives are included in off-gas samples collected on a daily basis in a nuclear power plant. Therefore, even if the concentration of the radioactivity in the off-gas sample rises and the occurrence of leakage is detected, the radioactivity of Kr-85m and Kr-85g in the nearby off-gas sample is measured and the intensity ratio is measured. If (Kr−85 g / m) is calculated, the burnup of the leaker can be obtained based on the correlation shown in FIG.

  In this way, by analyzing off-gas samples that are routinely measured, nuclides with a long half-life can be measured, so that the burnup of the leaker can be evaluated quickly and accurately when a leak occurs, particularly during plant operation. Furthermore, since the amount of radioactivity of Kr-85g is about 1/10 to 1/100 of Kr-85m, it can be sufficiently measured with a sample collected in a normal routine without requiring a large amount of sample. For this reason, according to this detection method, detection is possible even when the leaker is small and detection by reactor water Cs measurement is difficult, and the detection sensitivity is greatly improved.

  The radioactivity intensities of Kr-85m and Kr-85g can be obtained by measuring γ rays or β rays performed on an off-gas sample.

  FIG. 2 is a diagram showing the configuration of a damaged fuel assembly detection apparatus in which the radioactivity amount of Kr-85 g of an off-gas sample is obtained by γ-ray measurement.

As shown in FIG. 2, the present apparatus separates an off-gas sample 1 collected on a daily basis from a sample gas reservoir 3 filled with molecular sieves using N ₂ gas as a transfer gas (carrier gas) 2. After passing through the column 5 and removing only the Xe gas contained in the off-gas sample 1 by adsorbing to the column 5, Kr-85 g of 514 keV and Kr-85 m of γ-rays of the gas collected in the collection tube 6 are converted into Ge. The measurement is performed by the semiconductor detector 7, the spectrum is analyzed by the multichannel wave height analyzer 8, and the radioactivity intensities of both nuclides are compared.

  An off-gas sample collected by a normal routine contains Kr and Xe. The radioactivity intensity of the Kr-85g nuclide to be measured is much smaller than that of the Xe isotope. Therefore, it is difficult to directly measure the γ-rays of the collected sample with a Ge detector due to the interference by Xe radiation, but by using this apparatus, the radioactivity amount of Kr-85 g can be measured. And the radioactivity intensity | strength of both nuclides of obtained Kr-85g and Kr-85m obtained from routine measurement is compared, The burnup of a leaker can be calculated from this value.

  Next, a case where β-ray measurement is used (not shown) will be described. This is configured to obtain a radioactivity amount of Kr-85 g of an off-gas sample by β-ray measurement. In this case, a GM type βγ detector or a GM-NaI coincidence type β-ray scintillation detector is used as the radioactivity intensity measuring means. By measuring β-rays, it is possible to increase the detection sensitivity of Kr-85 g compared to the case of measuring γ-rays.

  The difference between the GM type βγ detector as the radiation detector and the GM-NaI coincidence type β ray scintillation detector is as follows. That is, since the GM type βγ detector measures both nuclides of Kr-85m and Kr-85g, when measuring Kr-85g, the off-gas sample is cooled in advance for 20 hours or more and Kr-85m having a short half-life. It is necessary to measure Kr-85 g after attenuating. On the other hand, in the case of a GM-NaI coincidence β-ray scintillation detector, only Kr-85g β-rays can be measured, so that it is not necessary to attenuate Kr-85m by cooling, and quick data collection is possible. Become.

  FIG. 3 is a graph showing the relationship between the irradiation days of the fuel assemblies loaded in the nuclear reactor and the radioactivity intensity ratio of the I isotope in the radioactive gas waste generated during the irradiation. FIG. 3 shows the relationship between the radioactivity intensity ratio of I-130 (I-130 / 133) to I-133 and the number of irradiation days (burning degree). The reason why these radioactivity intensity ratios are used in determining the burner burner will be described below.

  The reactor water I sample collected in a normal routine contains a plurality of types of I isotopes that are FPs of U and Pu and have different half-lives. Among these isotopes, although the half-lives of I-131 to I-135 are different from each other, they are all sufficiently shorter than the reactor operation period. Therefore, the radioactivity intensity of these nuclides does not depend on the burnup degree of the fuel, and, as in the case of Kr-85 m, becomes saturated as soon as combustion starts and takes almost a constant value throughout the combustion period.

On the other hand, since I-129 has a long half-life of 10 ⁷ years among I isotopes, it accumulates in the fuel rod due to combustion, and its radioactivity intensity gradually increases with the increase in combustion, but the change is directly Is hard to catch. However, I-130, which is a nuclide having a relatively short half-life, is generated by neutron capture reaction from I-129 accumulated and increased as the burnup increases. Therefore, the radioactivity intensity of I-130 can be regarded as reflecting the amount of production of I-129, and by knowing the radioactivity intensity of I-130, this can be used as an indicator of burnup. .

  As described above, in the nuclear reactor water, the radioactivity amount of any of the nuclides I-131 to I-135, for example I-133, which becomes a substantially constant value during the operation period is measured, while the burnup increases. The amount of radioactivity of I-130 accumulated and increasing is simultaneously measured to determine the ratio of both radioactivity intensity, and the burnup of the leaker is evaluated by using the relationship between the radioactivity intensity ratio and the burnup shown in FIG. It can be done.

  In addition, since the half-life of I-130 is as short as 12 hours, the detection sensitivity is increased as compared with the conventional case. The half-life of I-130 of 12 hours is also the length that allows samples to be taken and measured during the additional I release that occurs when the plant is shut down after the leak occurs. Therefore, according to this detection method, the burnup of the leaker can be evaluated even after the plant is stopped. The attenuation of I-130 due to the half-life between operation and sample measurement can be corrected by the change in the radioactivity intensity ratio of I-131 to I-135 in the same sample. The detection sensitivity of this detection method is more than 10 times that of the normal Cs method due to the increase in the concentration of I in the reactor water due to this additional release and by adopting the γ-ray spectrum analysis of the I isotope for the I-130 measurement. To rise.

  FIG. 4 is a graph showing the relationship between the number of irradiation days of the fuel assembly loaded in the nuclear reactor and the radioactivity intensity ratio of the I isotope in the radioactive gas waste generated during the irradiation. FIG. 4 shows the relationship between the radioactivity intensity ratio of I-128 (I-128 / 133) to I-133 and the number of irradiation days (burning degree). The reason why these radioactivity intensity ratios are used in determining the burner burner will be described below.

  As described above, the radioactivity intensity of I-131 to I-135 in the reactor water takes a substantially constant value throughout the combustion period regardless of the degree of combustion of the fuel. Although the half-lives are different, they are all sufficiently short compared to the reactor operation period. Therefore, the radioactivity intensity of these nuclides does not depend on the burnup degree of the fuel, and, as in the case of Kr-85 m, becomes saturated as soon as combustion starts and takes almost a constant value throughout the combustion period. Among these I isotopes, I-127 is focused.

  Like I-129, I-127 is a stable isotope of I, which is an FP of U or Pu, and performs a neutron capture reaction to produce I-128, a nuclide having a short half-life of 25 minutes. Therefore, the radioactivity intensity of I-128 reflects the amount of production of I-127, and knowing the radioactivity intensity of I-128 can be used as an index of burnup.

  As described above, in the nuclear reactor water, the radioactivity of one of the I-131 to I-135, for example, I-133, which becomes a substantially constant value during the operation period, is measured, and as the burnup increases, It is possible to simultaneously measure the amount of radioactivity of I-128 accumulated and increase to determine the ratio of both radioactivity intensity, and to evaluate the burnup of the leaker using the relationship between the radioactivity intensity ratio and the burnup shown in FIG. It can be done.

  Since the half-life of I-128 is as short as 25 minutes, not only the detection sensitivity is greatly increased compared to the conventional case, but also compared with the usual Cs analysis method that is performed when the plant is operating at the time of occurrence of the leaker, The detection sensitivity increases 10 times or more.

  Comparison of detection sensitivity between the above method and the conventional Cs analysis method will be described. In the nuclear power plant, when comparing the generated radioactivity intensity of Kr-85g, I-128, I-130, and Cs-137, Cs-137 is the largest, and then I-128 and I-130 are comparable. Kr-85g is the smallest. However, since the ratios of these nuclides released into the reactor water are different from each other, when comparing the radioactivity intensity of these nuclides actually released into the reactor water, Kr-85g is the largest, and then I-128. And I-130 are comparable, and Cs-137 is the smallest. At this concentration in the reactor water, I is about 10 times that of Cs. Therefore, in the method using I, the detection sensitivity is improved by about 10 times compared to the conventional Cs method. On the other hand, as with I, Kr can be measured with the amount of sample collected in a normal routine, so that the detection sensitivity can be improved to the same extent as the method using I compared with the conventional Cs method. .

<First Embodiment>
Prior to describing the detection device of the present invention, the sipper cap used in the present device will be described. FIG. 5 is a schematic view of a cross section of the present sipper cap.

  As shown in FIG. 5, the sipper cap 10 has a top portion 11 closely fixed at a predetermined position above the fuel gripping tool 30 at the tip of the mast 20 of the refueling machine, and the top of the fuel assembly 40 to be detected. By being attached to the portion 41, the fluid sample 13 can be isolated in the internal space 12. The top portion 11 of the sipper cap has a mast mounting portion 14 that is open so that the mast can be inserted, and a stopper 15 that is mounted to tightly fix the mast mounting portion 14 to the inserted mast outer periphery. Is provided. There is no restriction | limiting in particular in the shape of the stopper 15, For example, an attachment band etc. can be used satisfactorily.

  In the close fixing of the sipper cap 10 to the mast, first, the sipper cap 10 is arranged under the fuel gripping tool 30 at the tip of the mast 20, and then the sipper cap is inserted into the mast mounting portion 14 from the lower end. 10 is moved upward, and the stopper 15 is mounted so that the top 11 is closely fixed to a predetermined position above the fuel grip 30 of the mast 20. Alternatively, with the fuel gripping tool 30 inserted through the mast mounting portion 14 of the sipper cap 10, the fuel gripping tool 30 is attached to the mast 20, and the stopper 15 is attached so that the sipper cap 10 is firmly fixed to the mast 20. It may be.

  The internal space 12 of the sipper cap 10 has a shape that can accommodate the fuel grip 30 and the fuel handle 42 when the fuel grip 30 grips the fuel handle 42 of the target fuel assembly 40. A spring-type channel box holder 17 is provided at the lower edge portion 16 of the sipper cap 10 that is attached to the top of the fuel assembly 40 as a contact fixing means. The holder 17 has the inner periphery of the lower edge portion 16 of the sipper cap 10 adhered and fixed to the outer periphery 43 of the side wall surface of the fuel channel box of the fuel assembly 40 by the elasticity of the spring, and the isolated fluid sample is placed in the inner space. It has a structure that can store and hold. One end of the sample transfer pipe 50 is opened near the top 11 of the sipper cap, and the other end of the sample transfer pipe 50 is connected to a measuring means (not shown).

  As described above, since this sipper cap has a structure that is firmly fixed to the channel box, the reactor water existing between the fuel rods in the fuel assembly can be removed even while the fuel assembly is pulled up by the fuel exchanger. It can be isolated and stored inside and stored inside. Therefore, the radionuclide released from the leaker is sufficiently accumulated in the stored reactor water.

  The detection apparatus of the present invention comprises the sipper cap, a measuring means for detecting a fission product by measuring the radioactivity intensity of a sample isolated and collected by the sipper cap, and the sipper cap and the measuring means are connected. And a sample transfer pipe for transferring the fluid sample to the measuring means. The present detection apparatus can be applied to either a liquid sample or a gas sample, and the type of measurement means and the configuration of the sample transfer pipe can be appropriately selected according to the purpose.

<Second Embodiment>
Next, a method for detecting a damaged fuel assembly according to the present invention performed using the above-described detection device will be described. In collecting a fluid sample using the sipper cap 10, first, the sipper cap 10 fixed to the mast 20 of the fuel exchanger is lowered together with the fuel exchanger. Then, the fuel handle 42 of the fuel assembly upper tie plate is lowered until the fuel gripper 30 of the fuel changer grips it. When the fuel gripper 30 grips the fuel handle 42, the sipper cap lower edge 16 comes into close contact with the outer peripheral surface 43 of the side wall of the channel box, and the sipper cap 10 is integrated with the fuel assembly 40. Thereafter, when the mast 20 of the refueling machine is raised, the sipper cap 10 is more firmly fixed to the outer periphery 43 of the side wall surface of the channel box due to the resistance between the channel box and the reactor water. The fluid sample collected by the sipper cap 10 is transferred to a measurement means (not shown) through a sample transfer pipe 50 connected to the sipper cap 10, and the radioactivity intensity of the sample is measured by the measurement means. When the fission product is detected, the target fuel assembly is identified as a damaged fuel assembly.

  According to the detection method of the present invention described above, since the sipper cap can be attached to the target fuel assembly using the fuel exchanger, the sipper cap can be attached quickly and reliably. This method is suitable for in-core shipping. Furthermore, since the sipper cap according to the present invention has a structure in which the sipper cap is firmly attached to the fuel assembly, there is no dropout due to pressurization and gas leakage unlike the conventional sipper cap, and the fuel assembly portion can be sufficiently drained. Therefore, according to the detection method of the present invention, the target fuel assembly can be sufficiently isolated, so that the fuel temperature is easily raised and the sensitivity of the leaker detection is increased.

<Third Embodiment>
The detection apparatus according to the second aspect of the present invention will be described. This apparatus is provided with pressure adjusting means (not shown) for keeping the end of the measuring means (not shown) side of the sample transfer pipe 50 shown in FIG. 5 at a reduced pressure from the sipper cap side. This is the same as in the first embodiment, and a description of common parts is omitted. As the pressure adjusting means, for example, piping including a vacuum pump can be used.

<Fourth embodiment>
The detection method of the present invention using the apparatus shown in FIG. 5 corresponds to claim 8, and in this method, the pressure is stored in the internal space of the sipper cap due to the pressure difference inside and outside the sipper cap obtained by the pressure adjusting means. The fluid sample is transferred to the measuring means. The sample can be transferred easily and quickly.

<Fifth embodiment>
A detection apparatus according to the third aspect of the present invention will be described. FIG. 6 shows a gas sample as a fluid sample, that is, a rare gas component of FP released from a fuel rod into a reactor water as a leak occurs as a sample, and the present invention is configured to measure the radioactivity intensity. It is the schematic of the detection apparatus of one Embodiment of. This apparatus is configured to obtain a gas sample by providing a gas introduction pipe for introducing an isolation gas and a gas-liquid separation means for taking out the gas sample from the fluid sample to be transferred. Since the structure of the sipper cap 10 has already been described in FIG. 5, only its outline is shown in FIG. Moreover, since the method of attaching to the fuel assembly has already been described, the description thereof is omitted here.

  As shown in FIG. 6, a tank 51 serving as a gas-liquid separation unit is provided in the middle of the sample transfer pipe 50 connecting the sipper cap 10 and the measurement unit 70 in the present apparatus. That is, in the gas-liquid separation tank 51, one end of the first half of the sample transfer pipe 50 serving as the sample inlet from the sipper cap 10 and the second half of the sample transfer pipe 50 serving as the sample outlet for transferring the gas sample to the measuring means 70. One end of the part is open. Furthermore, at the tops of the sipper cap 10 and the gas-liquid separation tank 51, one ends of gas introduction pipes 60 for introducing an isolation gas are opened. A nitrogen gas supply source 62 as an isolating gas is connected to the other end of the gas introduction pipe 60. At the other end of the latter half of the sample transfer pipe 50, there are a sample chamber 52 which is a gas reservoir for storing a gas sample to be measured, and a pressure adjusting means for maintaining a reduced pressure between the sipper cap 10 and the measuring means 70. An air pump 53 as a main element is connected. The measuring means 70 includes a sample chamber 52 and a detector 71. By providing the sample chamber 52 in the measuring means 70, it is possible to prevent an increase in the background due to the radionuclide in the previous sample at the time of sample measurement, maintaining a good detection sensitivity and enabling a quick measurement. It becomes.

  Note that the rare gas separation column 54 shown by a dotted line in FIG. 6 is not used in this embodiment, and its bypass line is used in this embodiment. The rare gas separation column 54 will be described in the fourteenth embodiment.

<Sixth Embodiment>
A first detection method performed using the apparatus shown in FIG. 6 will be described. The present method corresponds to claim 9 and introduces a gas for isolation into the internal space of the sipper cap 10 through the gas introduction pipe 60 for a certain period of time, thereby providing an isolated furnace present in the space between the fuel rods. While aiming at the temperature rise of water, the gas radionuclide released by the fuel temperature rise is sampled with the reactor water through the sample transfer pipe 50 which is a gas outlet. Then, the liquid is separated from the collected gas-liquid mixed fluid using the gas-liquid separation means 51, the obtained sample gas, which is a radioactive noble gas, is introduced onto the gas chamber 52, and the radioactivity intensity is measured. To detect leica. In the measurement, the total radioactivity intensity of the sample gas was continuously measured with a GM type radiation detector.

  In the apparatus shown in FIG. 6, sampling and measurement are performed as follows. In the standby state, all the valves are closed in advance, and the valves V1, V7, V4 are opened prior to the sipper cap 10 being attached, the air pump 53 is operated, and the line after the gas-liquid separation tank 51 is in a reduced pressure state. After wearing the sipper cap 10 and opening V8, nitrogen as an isolating gas is introduced into the interior space of the sipper cap 10 for a predetermined time, then V1, V7 and V4 are tightened and V5 is opened to open the furnace from the fuel assembly. Water and gas are introduced into the gas-liquid separation tank 51 to separate the reactor water and the radioactive gas released from the damaged fuel rod and contained in the reactor water. The upper part of the gas-liquid separation tank 51 contains radioactive gas, and the lower part contains reactor water. Here, V5 is closed. Next, V <b> 2 of the gas introduction pipe 60 is opened, and nitrogen, which is an isolation gas, is introduced into the gas-liquid separation tank 51. Next, V4 and V7 are opened, nitrogen and all noble gases are transferred to the sample chamber 52, and the amount of radioactivity is measured by the GM type radiation detector 71. After the measurement is completed, V1 is opened, all gases are discharged, and one batch measurement is completed.

  In this method, nitrogen gas is used as the transfer gas, and the total radioactivity intensity of the sample gas is continuously measured by the GM type radiation detector to directly measure the total amount of radioactive gases Kr and Xe. The fastest and the best detection sensitivity. Therefore, it is effective when a fuel assembly having a relatively large damaged hole and speed is particularly required.

<Seventh embodiment>
A second detection method performed using the apparatus shown in FIG. 6 will be described. This method corresponds to claim 10 and is a sixth embodiment except that the total β-ray intensity of the sample gas is measured by using a simultaneous coefficient β-ray detector as a measuring means (claim 9). Corresponding to the above method), the damaged fuel assembly is detected. According to this method, since the total amount of β-rays of radioactive gases Kr and Xe is directly measured, the background is lowered and high detection sensitivity is obtained.

<Eighth Embodiment>
A third detection method performed using the apparatus shown in FIG. 6 will be described. This method corresponds to claim 13, except that the radioactivity intensity of the Xe isotope is measured by γ-ray spectrum analysis of the sample gas using a germanium semiconductor detector as the measuring means. In the same manner as in the method of the embodiment (corresponding to the method described in claim 9), the damaged fuel assembly is detected. According to this method, the background in the reactor water and I from the broken fuel rod can be distinguished, and high detection sensitivity can be obtained.

<Ninth embodiment>
The apparatus according to the ninth embodiment (corresponding to claim 4) is provided with a pipe having a rare gas separation column 54 in the latter half of the sample transfer pipe 50, as indicated by a dotted line in FIG. Others are the same as those of the fifth embodiment (corresponding to claim 3). The pipe removes Xe, which is a radioactive noble gas that interferes with measurement, through the rare gas separation column 54 from the gas obtained by separation in the gas-liquid separation tank 51, and uses the remainder as the gas sample to be measured. It is configured as follows.

<Tenth Embodiment>
A first detection method performed using the apparatus according to the ninth embodiment will be described. This method corresponds to claim 11, and the sampling and measurement of the sample are performed in the same manner as in the sixth embodiment (corresponding to the method described in claim 9) except that V3 is opened and closed instead of V7. Done.

  In this method, nitrogen gas is used as a transfer gas, a rare gas separation column is used, and the total radiation intensity can be measured to separate Kr and Xe, which are radioactive gases. Since the trace amount of Xe produced can be removed, the detection sensitivity can be increased.

<Eleventh embodiment>
A second detection method performed using the apparatus according to the ninth embodiment will be described. This method corresponds to claim 12 and is a tenth embodiment (explained in claim 11) except that the total β-ray intensity of the sample gas is measured using a simultaneous coefficient β-ray detector as a measuring means. Corresponding to the above method), the damaged fuel assembly is detected.

<Twelfth embodiment>
A detection apparatus according to the twelfth embodiment shown in FIG. 7 (corresponding to claim 5) will be described. This apparatus is configured for the purpose of detecting a damaged fuel assembly by isolating the fuel assembly using the sipper cap according to the present invention and measuring the radionuclide dissolved in the reactor water. Note that the structure of the sipper cap 10 and the manner in which the sipper cap 10 is attached to the fuel assembly has already been described, and a description thereof will be omitted here.

  As shown in FIG. 7, a liquid reservoir 56 is provided in the middle of the sample transfer pipe 50 that connects the sipper cap 10 and the measurement means 70 in the present apparatus. That is, the liquid reservoir 56 has one end of the first half portion of the sample transfer pipe 50 serving as the sample inlet from the sipper cap 10 and one end of the second half portion of the sample transfer pipe 50 serving as the sample outlet for transferring the gas sample to the measuring means 70. Is open. Connected to the liquid reservoir 56 is an air pump 53 which is a main element of pressure adjusting means for keeping the outside of the sipper cap at a reduced pressure from the inside. Further, the branched ends of the gas introduction pipe 60 are opened in the sipper cap 10 and the liquid reservoir 56, respectively. Connected to the other end of the gas introduction pipe 60 is an air compressor 61 for supplying air as a transfer gas to the sipper cap 10 or the liquid reservoir 56 as required. A sample chamber 52 for storing a liquid sample to be measured is connected to the other end of the latter half of the sample transfer pipe 50. Note that the measuring means 70 includes a sample chamber 52 and a detector 71.

<Thirteenth embodiment>
A first detection method performed using the apparatus according to the twelfth embodiment shown in FIG. 7 will be described. This method corresponds to claim 14, wherein the fuel body is isolated using the sipper cap 10, and the reactor water existing in the space of the isolated fuel assembly is dissolved radionuclide I released into the reactor water. They are collected together and introduced onto the sample chamber 52, and the radioactive intensity is measured to detect the leaker. In the measurement, the total radioactivity intensity of the sample liquid was continuously measured with a GM type radiation detector. In the apparatus shown in FIG. 7, sampling and measurement are performed as follows. In the standby state, all the valves are closed in advance, and the valve V5 is opened prior to the sipper cap 10 being attached, and the air pump 53 is operated to keep the line after the liquid reservoir 56 in a reduced pressure state. The sipper cap 10 is put on and V3 is opened, and the reactor water in the fuel assembly gap is introduced into the liquid reservoir 56. After this, V3 is closed, and the sipper cap 10 starts the next fuel body mounting operation. Next, the air compressor 61 is operated, V1, V2, and V4 are opened to introduce compressed air, whereby the reactor water in the liquid reservoir 56 is transferred to the sample chamber 52.

  According to this method, normal air using a compressor can be used as the transfer gas, and by directly filtering the reactor water, detection can be performed with a simple apparatus without requiring a gas-liquid separation means. Further, since it is not necessary to transfer the sample to the laboratory unlike the conventional method, the leaker can be detected quickly.

<Fourteenth embodiment>
A second detection method performed using the apparatus according to the twelfth embodiment shown in FIG. 7 will be described. This method corresponds to claim 15 and is the same as in the thirteenth embodiment except that the nuclear activity intensity of the sample liquid is measured by the Ge semiconductor detector. Perform detection. This method also has the same effect as the thirteenth embodiment.

<Fifteenth embodiment>
A third detection method performed using the apparatus according to the twelfth embodiment shown in FIG. 7 will be described. This method corresponds to claim 16 and is the same as in the thirteenth embodiment except that the nuclear type radioactivity intensity of the sample liquid is measured for each nuclear type by a Ge semiconductor detector and I is quantified. The fuel assembly is detected. This method also has the same effect as the thirteenth embodiment.

<Sixteenth Embodiment>
A detection apparatus according to the sixteenth embodiment shown in FIG. 8 (corresponding to claim 6) will be described.
This device isolates the fuel assembly using the sipper cap according to the present invention, collects the radionuclide I dissolved in the reactor water by adsorption, and measures its radioactivity intensity, thereby measuring the damaged fuel assembly. It is configured for the purpose of detecting the body. Note that the structure of the sipper cap 10 and the manner in which the sipper cap 10 is attached to the fuel assembly has already been described, and a description thereof will be omitted here.

  This apparatus is configured in the same manner as the apparatus shown in FIG. 7 except that an I adsorption filter 57 that adsorbs and collects I isotopes in the transferred liquid sample is provided instead of the sample chamber 52. .

<Seventeenth embodiment>
A first detection method performed using the apparatus according to the sixteenth embodiment shown in FIG. 8 will be described. This method corresponds to claim 17, and instead of transferring the reactor water in the liquid reservoir 56 to the sample chamber 52, the reactor water in the liquid reservoir 56 is returned to the pool through the I adsorption filter 57. The broken fuel assembly is detected in the same manner as in the thirteenth embodiment, except that the radioactivity of the I isotope adsorbed on 57 is measured. In the measurement, the total radioactivity intensity of the I isotope adsorbed on the filter 57 is continuously measured by a GM type radiation detector. This method also has the same effect as the thirteenth embodiment.

<Eighteenth embodiment>
A second detection method performed using the apparatus according to the sixteenth embodiment shown in FIG. 8 will be described. This method corresponds to claim 18 and is the same as in the seventeenth embodiment except that the nuclear type radioactivity intensity of the I isotope adsorbed on the filter 57 is measured by a Ge semiconductor detector. The damaged fuel assembly is detected.

  This method has the same effect as that of the thirteenth embodiment. As a further feature, since I in the reactor water is continuously collected and the isotope intensity ratio is measured, the influence of the background is reduced. This makes it possible to minimize the leak detection sensitivity and accuracy.

<Nineteenth embodiment>
A method for detecting a damaged fuel assembly corresponding to claim 22 will be described. This method uses a sipper cap according to the present invention to isolate a fuel assembly, and uses a device that combines both the device shown in FIG. 6 and the device shown in FIG. 7 or FIG. It measures both reactor water radioactivity.

  By measuring both, it is possible to cope with both the case where a large amount of I is released and the case where a rare gas is released depending on the vertical position of the fuel damaged assembly, and the detection accuracy and sensitivity can be improved. .

<20th Embodiment>
A method for detecting a damaged fuel assembly corresponding to claim 23 will be described. In the nineteenth embodiment, this method selects Kr-85 and I-131 as fission products to be measured.

<Twenty-first embodiment>
A method for detecting a damaged fuel assembly corresponding to claim 24 will be described. This method is used when the detection method according to the second embodiment or the fourth embodiment is carried out using the apparatus shown in FIG. 6, 7 or 8 using the sipper cap according to the present invention. In this method, after the fuel body is isolated, the amount of withdrawal of the fuel body is set to 3 m or more and the length of the fuel body by the fuel exchanger.

  By such drawing, the water depth of the fuel body becomes shallow, the pressure around it decreases, and the radioactive gas component accumulated inside the leaker is released. In addition, when the gas is inserted, the pressure of the inserted gas can be changed at the same time when the fuel assembly is pulled out, thereby increasing the fuel temperature and accelerating the release of the accumulated radioactive gas component. Since the entire fuel assembly is not pulled out, when the fuel assembly is inserted again to the original position, the operation for determining the insertion position can be omitted. Therefore, the fuel assembly can be pulled out and inserted quickly.

  Note that the release of radionuclides from the leaker depends greatly on temperature and pressure, and the increase in the temperature of the fuel rod and the decrease in pressure outside the fuel rod act to promote the release of the radionuclide in the fuel rod. Therefore, when this sipper cap is attached and the reactor water is isolated and the fuel assembly is extracted, the fuel rod external pressure can be lowered due to the extraction, and the temperature of the fuel rod can be increased sufficiently due to the isolation. Therefore, the release of radionuclide from the leaker fuel rod is promoted.

  Needless to say, this method is also applicable to the detection method according to claims 9 to 23.

The relationship between the irradiation days of the fuel assemblies loaded in the nuclear reactor and the Kr-85g activity intensity ratio (Kr-85g / m) to Kr-85m in the radioactive gas waste generated during irradiation is shown. Graph. Schematic which shows the structure of the detection apparatus of the broken fuel assembly which acquired the amount of radioactivity of Kr-85g of an off-gas sample by the gamma ray measurement. The graph which shows the relationship between the irradiation days of the fuel assembly loaded in the nuclear reactor, and the radioactivity intensity ratio (I-130 / 133) of the I isotope in the radioactive gas waste generated during irradiation. The graph which shows the relationship between the irradiation days of the fuel assembly loaded in the nuclear reactor, and the radioactivity intensity ratio (I-128 / 133) of the I isotope in the radioactive gas waste generated during irradiation. The schematic of the cross section of the sipper cap concerning this invention. The schematic diagram of the detection apparatus of one Embodiment of this invention comprised so that it might measure the radioactivity intensity | strength by making into a sample the FP rare gas component discharge | released from the fuel rod into the reactor water with generation | occurrence | production of a leaker. 1 is a schematic view of a detection apparatus according to an embodiment of the present invention configured to measure the radioactivity intensity of a radionuclide dissolved in reactor water. BRIEF DESCRIPTION OF THE DRAWINGS Schematic of the detection apparatus of one Embodiment of this invention comprised so that I which is a radionuclide melt | dissolved in reactor water was collected by adsorption | suction, and the radioactivity intensity | strength was measured. Schematic of the cross section of the conventional sipper cap.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 ... Off-gas sample, 2 ... Transfer gas, 3 ... Sample gas reservoir, 5 ... Noble gas separation column, 6 ... Collection tube, 7 ... Ge semiconductor detector, 8 ... Multichannel wave height analysis 10 ... Sipper cap, 11 ... Top, 12 ... Internal space, 13 ... Fluid sample, 14 ... Mast mounting part, 15 ... Stopper, 16 ... Lower edge, 20 ... Fuel changer 30 ... Fuel gripper, 40 ... Fuel assembly, 41 ... Top, 42 ... Fuel handle, 43 ... Outer periphery of fuel channel box side wall, 50 ... Sample transfer pipe, 51 ... ... Gas-liquid separation tank, 52 ... Sample chamber, 53 ... Air pump, 54 ... Rare gas separation column, 56 ... Liquid reservoir, 57 ... I adsorption filter, 60 ... Gas introduction pipe, 61 ... Air compressor, 62 …… Nitrogen Gas supply source, 70 ... measuring means, 71 ... detector, 100 ... conventional sipper cap, 200 ... fuel gripper, 300 ... fuel handle at the top of the fuel assembly, 400 ... sampling tube, 500 …… Air pipe

Claims (24)

A measuring means for detecting the fission product by measuring the radioactivity intensity of the sample,
A sipper cap that is attached to the top of the target fuel assembly of the detection operation and isolates the fluid sample in the internal space;
A device for detecting a damaged fuel assembly, comprising a sample transfer pipe for connecting the sipper cap and the measuring means to transfer an isolated fluid sample to the measuring means, and for identifying the damaged fuel assembly. And
The top portion of the sipper cap is closely fixed to a predetermined position above the fuel gripper of the fuel changer mast, and the internal space of the sipper cap is the fuel gripper and the target fuel assembly gripped by the fuel gripper. The sipper cap has a shape that can accommodate a fuel handle of the sipper cap, and the inner periphery of the lower edge portion of the sipper cap that is attached to the top of the target fuel assembly is disposed on the fuel channel side of the target fuel assembly. An apparatus for detecting a damaged fuel assembly, comprising: a fixing means for tightly fixing to an outer periphery of a wall surface and storing and holding the isolated fluid sample in the internal space.   Pressure adjusting means is further provided that keeps the measurement means side of the sample transfer pipe at a reduced pressure from the sipper cap side, and allows the fluid sample stored in the internal space to be transferred to the measurement means by an internal and external pressure difference. The apparatus for detecting a damaged fuel assembly according to claim 1, wherein:   A gas-liquid separation means further comprising a gas introduction pipe for introducing an isolation gas into the internal space of the sipper cap, and wherein the sample transfer pipe separates the liquid from the transferred fluid sample and takes out the gas sample. The apparatus for detecting a damaged fuel assembly according to claim 2, comprising:   The said sample transfer pipe | tube further has the interference noble gas removal means which removes interference noble gas from the gas taken out by the said gas-liquid separation means, and makes the remainder a gas sample. Damaged fuel assembly detection device.   The apparatus for detecting a damaged fuel assembly according to claim 2, wherein the fluid sample transferred after being stored in the internal space is a liquid sample.   6. The damaged fuel assembly according to claim 5, wherein the sample transfer pipe has I adsorption means for adsorbing and collecting I isotopes in the liquid sample to be transferred and separating the I isotope samples. Detection device. A sipper cap is attached to the top of the target fuel assembly of the detection operation to isolate the fluid sample in the inner space, and the radioactive intensity of the obtained sample is measured to detect the fission product and detect the damaged fuel assembly. A method for detecting a broken fuel assembly that identifies
The top of the sipper cap is closely fixed to a predetermined position above the fuel gripper of the refueling machine mast,
The internal space of the sipper cap has a shape capable of accommodating the fuel gripper and a fuel handle of a target fuel assembly gripped by the fuel gripper, and the sipper cap includes a target fuel assembly The inner periphery of the lower edge of the sipper cap attached to the top of the body is closely fixed to the outer periphery of the side wall of the fuel channel of the target fuel assembly so that the isolated fluid sample can be stored in the inner space. Closely fixing means are provided, and in isolating the fluid sample by attaching the sipper cap,
The fuel handle of the target fuel assembly is gripped by the fuel gripper to attach the sipper cap to the top of the target fuel assembly, and the inner periphery of the lower edge of the sipper cap is the fuel channel of the target fuel assembly. Closely fixed to the outer periphery of the side wall surface, the isolated fluid sample is stored in the internal space,
Next, the fluid sample is transferred to a measuring means through a sample transfer pipe, and the radioactive intensity of the fluid sample is measured by the measuring means to detect a fission product, and the damaged fuel assembly is identified. Method for detecting damaged fuel assemblies.   In transferring the fluid sample stored in the internal space to the measurement means via the sample transfer pipe, the measurement means side of the sample transfer pipe is kept at a reduced pressure from the sipper cap side and stored in the internal space. 8. The method for detecting a damaged fuel assembly according to claim 7, wherein the fluid sample is transferred to the measuring means by a pressure difference between inside and outside.   After storing the fluid sample in the internal space, an isolation gas is introduced, the gas is separated from the fluid sample to be transferred, the gas sample is taken out, and then the total radioactivity intensity of the taken out gas sample is measured. 9. The method for detecting a damaged fuel assembly according to claim 8, wherein the measurement is performed by means.   After the fluid sample is stored in the internal space, an isolating gas is introduced, the gas is separated from the fluid sample to be transferred, the gas sample is taken out, and then the total β-ray intensity of the taken out gas sample is measured. 9. The method for detecting a damaged fuel assembly according to claim 8, wherein the measurement is performed by means.   After storing the fluid sample in the internal space, an isolating gas is introduced, and after separating the gas from the fluid sample to be transferred, the interfering rare gas is removed and the gas sample is taken out, and the total radioactivity intensity of the gas sample is increased. 9. The method for detecting a damaged fuel assembly according to claim 8, wherein the measurement is performed by the measuring means.   After storing the fluid sample in the internal space, an isolating gas is introduced, and after separating the gas from the fluid sample to be transferred, the interfering noble gas is removed and the gas sample is taken out, and Kr-85 g of the gas sample is removed. 9. The method for detecting a damaged fuel assembly according to claim 8, wherein the β-ray intensity is measured by the measuring means.   After storing the fluid sample in the internal space, introducing an isolating gas, and measuring the nuclear type radioactivity intensity of the Xe isotope in the gas taken out from the fluid sample to be transferred by the measuring means. The method for detecting a damaged fuel assembly according to claim 8, wherein:   9. The method for detecting a damaged fuel assembly according to claim 8, wherein the total radioactivity intensity of the liquid sample transferred after being stored in the internal space is measured by the measuring means.   9. The method for detecting a damaged fuel assembly according to claim 8, wherein the nuclear type radioactivity intensity of the liquid sample stored after being stored in the internal space is measured by the measuring means.   9. The method for detecting a damaged fuel assembly according to claim 8, wherein the measuring means measures the radioactivity intensity of the I isotope of the liquid sample that is transferred after being stored in the internal space.   The I isotope is separated from the liquid sample stored in the internal space and then transferred by adsorption, and the total radioactivity intensity of the separated I isotope is measured by the measuring means. 9. A method for detecting a broken fuel assembly according to 8.   The I-isotope is separated by adsorption from a liquid sample that is stored in the internal space and then transferred, and the radioactivity intensity of the separated I-isotope is measured by the measuring means. Item 9. A method for detecting a damaged fuel assembly according to Item 8.   18. The method for detecting a broken fuel assembly according to claim 9, 11, 14, or 17, wherein the measuring means is a GM type radiation detector.   The method for detecting a damaged fuel assembly according to claim 10 or 12, wherein the measuring means is a coincidence β-ray detector.   19. The method for detecting a broken fuel assembly according to claim 13, 15, 16, or 18, wherein the measuring means is a Ge semiconductor detector.   8. The radioactivity intensity of the fission product in the gas sample transferred after being stored in the internal space and the radioactivity intensity of the fission product in the liquid sample are measured by the measuring means. 9. A method for detecting a damaged fuel assembly according to 8 and 8.   23. The method for detecting a damaged fuel assembly according to claim 22, wherein the fission product of the gas sample is Kr-85 g, and the fission product in the liquid sample is I-131.   While the isolated fluid sample is stored and held in the internal space, the fuel changer mast is moved up and down to pull out and insert the target fuel assembly from a predetermined position, and then the fluid sample is used for sample transfer. 9. The method for detecting a broken fuel assembly according to claim 7 or 8, wherein the fuel is transferred to a measuring means via a pipe.

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