JPH04268496A - Device and method for diagnostic abnormality of reactor - Google Patents

Abstract

PURPOSE:To detect quickly the leakage of a primary system coolant due to a break of a heat exchanger tube in a steam generator in a pressurized water reactor and thereby to diagnose the state of operation of the reactor. CONSTITUTION:A radiation detector 18 is provided for a main steam piping 8 of a secondary system through which steam generated from a steam generator 6 is made to flow into a turbine 9. By the radiation detector 18, the radiation is measured continuously as a dose rate or a counting rate. Measured data are transferred to an arithmetic processing system 20 and collated with references of normal values and the presense or absence of a leakage is diagnosed. Besides, radiation detectors 18 and 19 are provided on the upstream and downstream sides of the main steam piping 8 of the secondary system respectively and the data are processed as a variation with time in the arithmetic processing system 20. By this constitution, the scale of a broken part of a heat exchange tube 7 in the steam generator 6 is diagnosed on the basis of the value of the radioactive concentration in the main steam on the occasion of the leakage.

Description

[Detailed description of the invention]

[Object of the invention]

0002

[Industrial Application Field] The present invention relates to a pressurized water nuclear reactor (hereinafter referred to as
The present invention relates to a nuclear reactor abnormality diagnosis device and its diagnosis method for diagnosing leakage of primary coolant from heat transfer tubes in a steam generator in a nuclear reactor (referred to as PWR).

0003

2. Description of the Related Art A conventional pressurized water nuclear power plant will be schematically explained with reference to FIG. In the figure, reference numeral 1 indicates a reactor building, 2 indicates a reactor containment vessel, and 3 indicates a reactor vessel. The primary coolant heated in the reactor core 4 in the reactor vessel 3 flows through the primary system piping 5 and flows into the heat transfer tube 7 in the steam generator 6 . The secondary coolant in the steam generator 6 exchanges heat with the heat exchanger tubes 7 and is heated to become high-temperature steam, which flows through the secondary main steam piping 8 and flows into the turbine 9. The turbine 9 rotates and drives the generator 10 to generate electricity. The steam that has completed its work in the turbine 9 flows into the condenser 11, where it is cooled and becomes condensed water. This condensate flows through the secondary system water supply pipe 12 and is supplied to the steam generator 6 as a secondary system coolant. On the other hand, the primary coolant flowing through the heat transfer tubes 7 in the steam generator 6 is cooled by exchanging heat with the secondary coolant, and is cooled from the primary main cooling pipe 13 into the reactor vessel 3 by the pump 14. It is heated in the reactor core 4. In addition,
In the figure, 15 is a control rod, 16 is a pressurizer, and 17 is a spray pipe. In addition, 22 is an exhaust gas system bleed pipe, 2
3 is a turbine exhaust gas system monitor that measures radioactive substances in the exhaust gas.

By the way, in the steam generator 6, a large number of heat transfer tubes 7 made of inverted U-shaped thin tubes are attached to a tube plate, and all of these heat transfer tubes 7 are inspected during regular inspections. The soundness has been confirmed. If the heat transfer tubes 7 are damaged during reactor operation and primary coolant leaks, the leaked radioactive materials will pass through the secondary main steam piping 8 and move to the turbine 9, and then to the condenser 11. It will return to the water supply system through. In PWR, some steam (gas) from condenser 11
The exhaust gas monitor 23 is configured to issue an alarm if the extracted steam contains radioactive substances exceeding a set alarm value.

[0005]

[Problems to be Solved by the Invention] In the conventional PWR, when the primary coolant leaks, an alarm is issued when the amount of radioactive material leaked exceeds the alarm setting value of the exhaust gas monitor 23. However, as shown in FIG. 1, the exhaust gas monitor 23 is located in the exhaust gas system bleed pipe 22 led out from the condenser 11 at the terminal end even in the secondary system. Short-term radionuclides such as N16 (half-life: 7 seconds) and C15 (half-life: 2 seconds), which are the main source nuclides of the primary system coolant, leak into the secondary system due to damage to the heat transfer tubes 7 in the steam generator 6. The half-life nuclide decays until it reaches the exhaust gas monitor 23 and is hardly measured. Therefore, the radionuclides targeted by the exhaust gas monitor 23 due to leakage of the primary coolant are rare gases, corrosion products, nuclear fission products, and the like. However, the amount of these radionuclides present in the primary coolant is smaller than that of N16 and C15, and since the part of the radioactive nuclides that leaked into the main steam is extracted and measured, it is possible that a trace amount of leakage may occur. It becomes difficult to detect. In this way, the device is unable to detect any signs of abnormality due to trace leakage. Additionally, there are problems such as the inability to grasp the concentration of leaked radioactive materials and abnormal conditions.

The present invention has been made to solve the above-mentioned problems, and is a nuclear reactor that can quickly detect a small amount of leakage of primary coolant from a heat transfer tube in a steam generator in a pressurized water reactor. An object of the present invention is to provide an abnormality diagnosis device and a diagnosis method thereof. [Structure of the invention]

[0007]

[Means for Solving the Problems] In the first aspect of the present invention, a pressurizer and a steam generator are connected to a pressurized water reactor vessel via a primary system piping, and from the secondary system outlet of the steam generator to the turbine. It is characterized in that a secondary main steam piping is connected, a radiation detector is connected to the secondary main steam piping, and a calculation processing system is connected to the radiation detector.

[0008] In the second aspect of the present invention, a pressurizer and a steam generator are connected to a pressurized water reactor vessel via a primary system piping, and a secondary system is connected from a secondary system outlet of the steam generator to a turbine. At least two radiation detectors are installed as measurement points in the main steam piping, and the steam flow rate is calculated from the attenuation rate of each dose rate measured using short half-life nuclides detected by these radiation detectors as source nuclides, The method is characterized in that the radioactivity concentration at the measurement point is calculated by back-calculating up to the secondary system outlet of the steam generator.

[0009]

[Operation] In the abnormality diagnosis device according to the first invention, the radiation detector is installed on the upstream side of the flow path of the secondary main steam piping (at a position close to the steam generator), and the
By detecting radioactivity before it decays, it becomes possible to diagnose early signs associated with the leakage of minute amounts of radioactive material. This means that the primary coolant that leaks into the main steam due to heat exchanger tube breakage in the steam generator contains radioactive rare gases, corrosion products, nuclear fission products, etc.
It contains N16, C15, etc., but the amount of N16 and C15 contained per unit volume of primary system coolant during normal reactor operation is the largest, and because it easily transfers to the main steam system, early detection is possible. Next, since the dose rate from the secondary main steam piping is proportional to the radioactivity concentration in that piping, the dose rate at the measurement point is determined at an arbitrary concentration by calculation in advance, and the dose rate obtained by the radiation detector is calculated in advance. Ratio between the measured value of the dose rate and the calculated value of the dose rate calculated by the above calculation (actual value/calculated value)
is normalized to the concentration used in the calculation, and a calculation processing system calculates the radioactivity concentration for the dose rate of the actual measurement value.

[0010] In the abnormality diagnosis method in the second aspect of the invention, self-diagnosis is first performed on the data obtained by measuring with radiation detectors at a plurality of locations. This means that if a leak occurs, the measured values will change in the order of upstream and downstream along the main steam flow path, and the measured values will differ by the difference in arrival time, but radiation detection If it is a device failure or noise, it can be determined because it only appears in one measurement result. Next, if the value is significantly different from the normal value,
It is determined that there is a leak, and the arrival time of the steam from the upstream measurement point to the downstream measurement point can be calculated from the difference in measured values (attenuation rate). A method for estimating the amount of leaked radioactivity is shown.

[0011]N/N0 = e(-λt)t=ln(N/
N0 )/(-λ) Here, N0 is the upstream measured value, N is the downstream measured value, and λ
is a constant and is determined by the nuclide. t is the arrival time of steam from the upstream measuring point to the downstream measuring point. Furthermore, if the length of the flow path between the measurement points is known, the flow velocity S (m/sec) of steam can be calculated using the following formula. S=l/t (m/sec) where l is the flow path length (m) between measurement points, and t is the arrival time (sec). Based on the calculated steam flow velocity and the piping distance from the measurement point to the secondary system outlet of the steam generator, the arrival time of the steam is determined in the same way as above, and the concentration calculated earlier is calculated backwards to reach the main steam generator. The amount of leaked radioactivity at the secondary system outlet can be estimated.

0012

[Example] First and second embodiments of the present invention with reference to the drawings
An example will be explained. FIG. 1 schematically shows a pressurized water nuclear power plant including an embodiment of the present invention. In the figure, code 1
2 indicates the reactor building, 2 indicates the reactor containment vessel, and 3 indicates the reactor vessel. The primary coolant heated in the reactor core 4 in the reactor vessel 3 flows through the primary system piping 5 to the steam generator 6.
It flows into a large number of heat exchanger tubes 7 inside. The secondary coolant in the steam generator 6 exchanges heat with the heat exchanger tubes 7 and is heated to become high-temperature steam, which flows through the secondary main steam piping 8 and flows into the turbine 9. The turbine 9 rotates and drives the generator 10 to generate electricity. The steam that has completed its work in the turbine 9 flows into the condenser 11, where it is cooled and becomes condensed water. This condensate is transferred to the secondary system water supply pipe 12
The water is supplied to the steam generator 6 as a secondary coolant. On the other hand, the primary coolant flowing through the heat transfer tubes 7 in the steam generator 6 is cooled by exchanging heat with the secondary coolant, and is cooled from the primary main cooling pipe 13 into the reactor vessel 3 by the pump 14. It is heated in the reactor core 4. In addition, 15 in the figure is a control rod, 16
1 indicates a pressurizer, and 17 indicates a spray tube.

In the first embodiment (first invention), in order to measure the radioactivity in the steam in the secondary main steam pipe 8, a measurement point A is set facing the pipe 8. One radiation detector 18 is provided. This first radiation detector 1
8 is connected to the arithmetic processing system 20 by a signal cable 21.

The position of the first radiation detector 18 in the first embodiment is preferably close to the outlet 6a of the steam generator 6. The radiation detector 18 uses an ionization chamber type or a plastic scintillation type, and detects MCS as the dose rate or count rate.
(Multi-channel scaling) mode for continuous measurement. Measured data is transferred in real time via a signal cable 21 such as a coaxial cable or an optical cable to an arithmetic processing system 20 comprising an arithmetic processing device (such as a personal computer) equipped with data input/output and arithmetic functions. This arithmetic processing system 20 searches for peaks using smoothed second derivatives, etc., and determines the presence or absence of leakage based on whether the statistical fluctuation exceeds three times (3σ) of the normal value, and determines the dose rate. Convert it to radioactivity concentration, output the value and amount of change, and predict the leakage transition based on the results. In addition, by determining in advance the radioactivity concentration of the radionuclide to be measured in the primary coolant and the transfer rate from the steam generator 6 to the secondary main steam piping 8, it is possible to predict the concentration of the radionuclide in the main steam in the event of a leak. The scale of the damaged portion of the heat exchanger tube 7 can be estimated from the value of the radioactivity concentration.

According to the first embodiment, the secondary main steam pipe 8 connecting the steam generator 6 and the turbine 9 is facing the upstream side (toward the steam generator) along the main steam flow path. A dose rate measurement system with a radiation detector 18 installed at a nearby location) and the data obtained from this measurement system are compared with normal conditions to determine whether there are abnormal signs due to leakage and to measure from the dose rate. Calculate the radioactivity concentration in the main steam at the point,
It consists of an arithmetic processing system that can process changes over time. If the heat transfer tube 7 in the steam generator 6 is damaged, the radioactive substances in the primary coolant will reach the radiation detector 18 in a few seconds. Note that since the radioactivity concentration in the main steam under normal conditions is originally equal to zero, the output of the detector is expected to change significantly. Especially N16 and C1
When 5 is used as a radiation source, the energy of the emitted gamma rays is high (6 MeV, 5 MeV), so even a minute amount can be sufficiently detected even outside the secondary main steam pipe (wall thickness 2 to 3 cm) 8. can.

Next, a second embodiment of the present invention (second invention)
will be explained with reference to FIGS. 1 and 2. In addition, Figure 1
Since this is almost the same as the explanation of the first embodiment, only the main components will be explained. In the second embodiment of the present invention, A and B in FIG. 1 are measurement points on the upstream side and downstream side of the secondary main steam piping 8, respectively, and the first and second radiation detection stations are installed at those locations. Install containers 18 and 19. These radiation detectors 18 and 19 are connected to a signal cable 2 to an arithmetic processing system 20.
1, and using an ion chamber type or plastic scintillation type, etc., the dose rate or count rate is M
Measure continuously in CS (multichannel scaling) mode. If the heat transfer tube 7 in the steam generator 6 is damaged and leakage of the primary coolant occurs, the measured values will change in the order of A and B, and the measured values will differ by the difference in arrival time. Therefore, if there is a failure or noise in the radiation detector, it should appear in only one measurement result. Said,
After performing self-diagnosis on the measured values in the arithmetic processing system 20,
The peak is searched for by smoothing second-order differential, etc., and one judgment criterion is whether the statistical variation exceeds three times (3σ) of the normal value. If it is determined that there is a sign of significant leakage, the measured value is immediately converted to radioactivity concentration using the dose rate and radioactivity concentration conversion coefficient calculated in advance, and the attenuation at measurement points A and B is further calculated. The main steam flow rate is calculated from the rate, the arrival time is determined from the piping distance from the measurement point to the secondary system outlet 6a of the steam generator 6, and the radioactivity concentration of A or B is calculated backward in time to determine the steam flow rate. The radioactivity concentration leaked from the primary system at the secondary system outlet 6a of the generator 6 is calculated. By processing the above-mentioned series of data as changes over time in the processing system, the transition of abnormality can be predicted immediately. By determining the transfer rate to the main steam system, it is possible to estimate the scale of the damaged part from the radioactivity concentration in the main steam at the time of leakage. By the way, let's estimate the amount of leakage that can be diagnosed using the following assumptions. The N16 radioactivity concentration in the primary coolant in the heat exchanger tube 7 was set to 40 (μCi/g).
If the main steam flow rate in the secondary system is 1700 (T/h) and a leak of L (g/s) occurs from the heat exchanger tube 7, the leakage amount per hour is 144 L (mCi/h). Become. If 50% of this is transferred to the main steam flow rate of 1500 T/h, approximately L x 5 x 10-5 (μCi/g) will exist in the main steam. Here, according to calculations, when there is 1 (μCi/g) of N16 in the main steam pipe, the dose rate at the surface of the main steam pipe (650A) is approximately 30
(mR/h), so L×5×10-5(μC
i/g), L x 1 x 10-3 (mR
/h), and the lower measurement limit of the radiation detector is 0.1 (m
R/h), the leakage rate L is 100 (g/s). This amount is the primary coolant flow rate per main steam generator unit of 15000 (T/h), and the flow rate per heat exchanger tube when the heat exchanger tubes 7 in the steam generator 6 are 3382 per unit. It is about 0.08% of 1200 (g/s).

According to the second embodiment (second invention), the steam pipe is located on the upstream side (steam A dose rate measurement system is installed with a plurality of radiation detectors 18 and 19 spaced apart (a position close to the generator) and a downstream side (a position near the turbine), and a plurality of data obtained from this dose rate measurement system are installed. Based on this, self-diagnosis is performed to determine whether the signal is noise or a significant signal, and a comparison is made with normal conditions to determine whether there are abnormal signs due to leakage, and the radioactivity concentration in the main steam at each measurement point is calculated from the dose rate. , and further from the secondary system outlet 6a of the steam generator 6 to the measurement points A and B based on the plurality of dose rates obtained along the flow path.
Find the main steam arrival time. Therefore, calculation of the leaked radioactivity concentration at the secondary system outlet 6a of the steam generator 6 can be diagnosed based on changes over time.

[0018]

According to the present invention, even if even a small amount of primary coolant leaks into the secondary coolant from the heat exchanger tubes in the steam generator, the dose rate of the measurement system will be reduced immediately because it is close to the point of occurrence. When a change appears in , it is possible to diagnose the early signs of an abnormal phenomenon, and by processing this data as a change over time, it is possible to predict the course of the abnormality and to diagnose the state of the abnormal location.

[Brief explanation of the drawing]

FIG. 1 is a block diagram schematically showing a pressurized water nuclear power plant including an embodiment of the present invention.

FIG. 2 is a block diagram of a measurement system and an arithmetic processing system in the present invention.

[Explanation of symbols]

1...Reactor building, 2...Reactor containment vessel, 3...Reactor vessel, 4...Reactor core, 5...Primary system piping, 6...Steam generator, 7...Heat transfer tube, 8...Secondary system main steam piping, 9 ... Turbine, 10 ... Generator, 11 ... Condenser, 12 ... Secondary system water supply pipe, 13 ... Primary system main cooling pipe, 14 ... Pump, 15 ... Control rod, 16 ... Pressurizer, 17 ... Spray pipe, 18 ...first radiation detector,
19... Second radiation detector, 20... Arithmetic processing system, 21...
Signal cable, 22...Exhaust gas system extraction piping, 23...Turbine exhaust gas system monitor.

Claims (2)

[Claims] Claim 1: A pressurizer and a steam generator are connected to a pressurized water reactor vessel via a primary system piping, and a secondary system main steam piping is connected from a secondary system outlet of the steam generator to a turbine, A nuclear reactor abnormality diagnosis device characterized in that a radiation detector is connected to the secondary main steam piping, and a calculation processing system is connected to the radiation detector. [Claim 2] A pressurizer and a steam generator are connected to the pressurized water reactor vessel via a primary system piping, and a secondary system main steam piping is connected from the secondary system outlet of the steam generator to the turbine. At least two radiation detectors are provided as measurement points, and the vapor flow rate is calculated from the attenuation rate of each dose rate measured using short half-life nuclides detected by these radiation detectors as source nuclides. A nuclear reactor abnormality diagnosis method characterized by calculating the radioactivity concentration back to the secondary system exit of the steam generator.