JP2008275358A - Device and method for detecting damage in nuclear reactor - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To detect quickly a damage in a nuclear reactor during a normal operation of the reactor. <P>SOLUTION: This detection device 10 of a damage in the reactor for detecting a damage in the reactor during a normal operation of a boiling water reactor has a measuring part 12 for measuring a physical quantity in the reactor during the normal operation of the boiling water reactor; a database part 13 for storing beforehand a fluctuation pattern of a physical quantity in the reactor, when, for example, a damage exists in a core shroud 11 in the boiling water reactor; and a comparison determination part 14 for comparing a fluctuation pattern of a physical quantity in the boiling water reactor measured by the measuring part with the pertinent fluctuation pattern of the physical quantity stored in the database part, determining whether the fluctuation patterns agree with each other or not, and determining existence of a damage in the reactor. <P>COPYRIGHT: (C)2009,JPO&INPIT

Description

  The present invention relates to an in-reactor damage detection apparatus that detects damage in the reactor during normal operation of the reactor.

  Many boiling water nuclear power plants (hereinafter abbreviated as BWRs) in Japan have been operating for a long period of more than 20 years, and replacement work for the core shroud is being promoted as a countermeasure against the deterioration over time. However, when power increase operation, long-term cycle operation, etc. are introduced in the future, it is expected that the heat output in the furnace, the coolant flow rate, the neutron irradiation amount, etc. will become more severe than the present situation.

The failure of the core shroud exposed to such a severe environment, in particular, the non-conformity event that the core shroud is subject to penetration failure such as stress corrosion cracking (SCC) has been attracting attention from the past. The thing is proposed by the following patent documents 1-6. In these patent documents, Patent Documents 1 to 3 are related to measures for preventing damage to the core shroud, and Patent Documents 4 to 6 are related to a method for identifying or repairing a damaged part at the time of stopping inspection (periodic regular inspection). is there.
Japanese Patent Laid-Open No. 3-235096 JP 2004-233258 A JP 2005-69732 A JP 2005-308557 A JP 2005-308459 A JP-A-7-248397

  Conventionally, the soundness of the core shroud has been confirmed by various inspection operations in regular inspections, but there is an increasing need to monitor the soundness of the reactor even during normal operation of the reactor. In other words, if damage such as through cracks occurs in the core shroud during normal operation of the reactor, it is possible to quickly identify and detect this damaged part during normal operation of the reactor from the viewpoint of microsymptom maintenance / subsequent maintenance. is necessary. However, there is no such means at present.

  An object of the present invention has been made in view of the above circumstances, and provides an in-reactor damage detection apparatus and method that can quickly detect damage in the reactor during normal operation of the reactor. is there.

  Another object of the present invention is to provide an in-reactor damage detection apparatus and method that can quickly identify and detect a damage site in a reactor during normal operation of the reactor.

  The in-reactor breakage detection apparatus according to the present invention is an in-reactor breakage detection apparatus that detects breakage in the reactor during normal operation of the reactor, and the physical quantity in the reactor is determined during normal operation of the reactor. An actual measurement unit to be measured, a database unit for preliminarily storing a fluctuation pattern of a physical quantity in the nuclear reactor when damage is present or absent in the nuclear reactor, and a physical quantity in the nuclear reactor measured by the actual measurement unit Comparison of the fluctuation pattern of the corresponding physical quantity stored in the database unit, and whether or not these fluctuation patterns match to determine whether or not there is damage in the reactor And a determination unit.

  Further, the in-reactor breakage detection apparatus according to the present invention is an in-reactor breakage detection apparatus that detects breakage in the reactor during normal operation of the reactor, and is in the reactor during normal operation of the reactor. An actual measurement unit for measuring a physical quantity, a numerical analysis unit for obtaining a fluctuation pattern by a numerical analysis when the physical quantity measured by the actual measurement unit is assumed to be free from damage in the nuclear reactor, and the actual measurement unit The measured fluctuation pattern of the physical quantity in the reactor is compared with the fluctuation pattern of the corresponding physical quantity obtained by the numerical analysis unit, and it is determined whether or not these fluctuation patterns match, And a comparison / determination unit for determining whether or not there is damage.

  Furthermore, the in-reactor breakage detection method according to the present invention is an in-reactor breakage detection method for detecting breakage in the reactor during normal operation of the reactor, wherein there is breakage in the reactor, or The fluctuation pattern of the physical quantity in the reactor when it does not exist is stored in the database unit in advance, the fluctuation pattern of the physical quantity measured in the reactor during the normal operation of the reactor, and the corresponding physical quantity stored in the database unit It is characterized by comparing the fluctuation patterns, judging whether or not these fluctuation patterns match, determining the presence or absence of damage in the reactor, and detecting the damage in the reactor .

  Further, the in-reactor damage detection method according to the present invention is an in-reactor damage detection method for detecting damage in the reactor at the time of normal operation of the reactor, and in the reactor at the time of normal operation of the reactor. The physical quantity is measured, and the fluctuation pattern when the measured physical quantity is assumed to be free of damage in the nuclear reactor is obtained by numerical analysis, and the fluctuation pattern of the measured physical quantity described above and the corresponding obtained from the numerical analysis are obtained. Comparing with fluctuation patterns of physical quantities, judging whether these fluctuation patterns match, determining whether there is damage in the reactor, and detecting damage in the reactor It is.

  According to the in-reactor damage detection apparatus and method according to the present invention, the fluctuation pattern of the physical quantity measured in the reactor during normal operation of the reactor, and the damage in the reactor stored in advance in the database unit are detected. Compare the fluctuation pattern of the physical quantity in the reactor when it exists or does not exist, determine whether these fluctuation patterns match, and determine whether there is damage in the reactor. During normal operation of the reactor, damage in the reactor can be detected quickly.

  Further, according to the in-reactor damage detection apparatus and method according to the present invention, there is no damage in the reactor with respect to the fluctuation pattern of the physical quantity measured in the reactor during normal operation of the reactor and the measured physical quantity. Comparison of fluctuation patterns obtained by numerical analysis with the assumption that the fluctuation patterns match, and whether these fluctuation patterns match, and whether or not there is damage in the reactor, During operation, damage in the reactor can be detected quickly.

  The best mode for carrying out the present invention will be described below with reference to the drawings.

[A] First embodiment (FIGS. 1 to 4)
FIG. 1 is a block diagram showing a first embodiment of an in-reactor damage detection apparatus according to the present invention. FIG. 2 is a cross-sectional view showing a boiling water reactor for explaining physical quantities to be measured in the in-reactor damage detection apparatus of FIG.

  The in-reactor breakage detection apparatus 10 in the present embodiment detects breakage of the core shroud 11 as the reactor internal structure during normal operation of the reactor, and further identifies and detects the breakage site of the core shroud 11. The apparatus includes an actual measurement unit 12, a database unit 13, a comparison determination unit 14, and a display unit 15.

  The actual measurement unit 12 measures a physical quantity in the nuclear reactor during normal operation of the nuclear reactor. When the nuclear reactor is the boiling water reactor 16 shown in FIG. 2, the actual measurement unit 12 has physical quantities such as LPRM (local neutron flux), APRM (core average neutron flux), core flow rate, core differential pressure, and JP (jet pump). At least one of a flow rate, a PLR (recirculation) pump flow rate, a PLR pump speed, a reactor dome pressure, a bottom drain temperature of the in-reactor coolant, a shroud internal / external differential pressure, and a in-reactor coolant temperature difference is measured.

  Here, the boiling water reactor 16 accommodates a core 18 in a reactor pressure vessel 17, and a large number of fuel assemblies (not shown) constituting the core 18 are surrounded by the core shroud 11 and support the core. Supported by the plate 20 and the upper grid plate 21. The upper part of the core shroud 11 is closed by a shroud head 23, and a steam / water separator 24 is installed in the shroud head 22 via a stand pipe 23. In the reactor pressure vessel 17, a steam dryer 25 is disposed above the steam separator 24.

  The steam generated in the core 18 is separated from the water by the steam separator 24, dried by the steam dryer 25, reaches the upper dome 26, passes through the main steam system from the main steam nozzle 27, and the turbine system ( (Both not shown). Steam that has worked in the turbine system becomes condensate, and is supplied as a coolant into the reactor pressure vessel 17 through a water supply system (not shown). The coolant is pressurized by the recirculation pump 29 of the reactor recirculation system 28, and a plurality of jet pumps 30 installed in the downcomer portion 32 between the reactor pressure vessel 17 and the core shroud 11 are used for the core 18. It is led to the lower core lower plenum 31.

  Note that the downcomer portion 32 indicates a range from the annular portion between the reactor pressure vessel 17 and the core shroud 11 to the portion between the reactor pressure vessel 17, the stand pipe 23 and the steam / water separator 24.

  The LPRM is a value measured by a plurality of neutron flux monitors installed in the axial direction and the circumferential direction of the core 18, and APRM is an average value of the values of the plurality of LPRMs. The core flow rate is the flow rate of the coolant flowing in the channel box of each fuel assembly in the core 18. Further, the core differential pressure is a pressure difference between the core 18 and the lower core plenum 31 and is generally proportional to the square of the core flow rate.

  The JP flow rate is the discharge flow rate of the jet pump 30. The PLR pump flow rate is the coolant flow rate that flows through the reactor recirculation system 28 by the recirculation pump 29, and the PLR pump speed is the rotational speed of the recirculation pump 29. Further, the reactor dome pressure is the pressure in the upper dome 26 of the reactor pressure vessel 17, and the bottom drain temperature of the reactor coolant is the coolant temperature of the reactor core lower plenum 31. Furthermore, the differential pressure inside and outside the shroud is a pressure difference between the core 18 and the downcomer portion 32, and the temperature difference between the coolants in the reactor is a temperature difference of the coolant between the core 18 and the downcomer portion 32.

  The database unit 13 shown in FIG. 1 organizes the fluctuation pattern (fluctuation tendency) of physical quantities in a nuclear reactor, for example, the boiling water reactor 16 in FIG. Stored in advance. For example, FIG. 3 shows main physical quantities (APRM, core flow rate, core differential pressure, reactor dome pressure, PLR pump speed, PLR pump flow rate) when the core shroud 11 in the reactor pressure vessel 17 is damaged. The fluctuation pattern is arranged for each damaged part. In this case, the damaged portion of the core shroud 11 includes an upper shroud portion (mainly a weld line H1 portion), a middle shroud portion (mainly a weld line H4 portion), and a lower shroud portion (mainly a weld line H7 portion).

  Next, the fluctuation pattern of the main physical quantity for each damaged part of the core shroud 11 will be described.

  When the upper shroud portion (mainly the weld line H1 portion) is damaged, a part of the coolant in the core upper plenum 33 surrounded by the shroud head 22 is removed from the damaged portion (for example, crack) of the core shroud 11. It flows out to the upper part of the downcomer 32 outside the stand pipe 23 and the steam separator 24. However, this flowing out coolant is in a two-phase state of steam and water, of which steam accounts for 70%, so that the pressure loss increases and the amount of outflow at this time becomes very small.

  At this time, since the pressure of the core upper plenum 33 slightly decreases due to the outflow from the damaged portion of the coolant, the coolant easily flows in the fuel assembly of the core 18, the core flow rate and the core differential pressure slightly increase, The flow rate (in-channel flow rate) in the channel box of the fuel assembly also slightly increases. Thus, when the in-channel flow rate of the coolant is increased, the thermal neutron flux is increased and the APRM is increased.

  On the other hand, a part of the coolant in the two-phase state in the core upper plenum 33 flows out from the damaged portion of the core shroud 11 to the outside of the stand pipe 23, so that the two-phase rising in the steam / water separator 24 is generated. The coolant in the state is reduced, and the amount of steam flowing from the steam separator 24 to the steam dryer 25 is reduced. For this reason, the reactor dome pressure of the upper dome 26 in the reactor pressure vessel 17 tends to decrease. Further, when the reactor recirculation system 28 is operated in the manual mode, the PLR pump speed and the PLR pump flow rate do not change.

  When the middle shroud portion (mainly the weld line H4 portion) is damaged, a part of the coolant from the gap (fuel bypass portion) between the fuel assemblies of the core 18 is downcomed from the damaged portion of the core shroud 11. It flows out to the part 32. However, this flowing out coolant is in a two-phase state of steam and water, of which steam accounts for 70%, so that the pressure loss increases and the amount of outflow at this time becomes very small.

  At this time, since the pressure in the fuel bypass corresponding to the middle shroud portion is slightly reduced due to the outflow from the damaged portion of the coolant, the coolant easily flows in the gaps between the fuel assemblies and in the fuel assemblies. The flow rate and core differential pressure increase slightly. As the in-channel flow rate through which the coolant flows through the fuel assembly increases, the thermal neutron flux increases and APRM increases.

  As the thermal neutron flux increases and the fission reaction in the fuel assembly is promoted, the amount of steam generated increases. The steam is introduced into the upper dome 26 of the reactor pressure vessel 17 through the stand pipe 23, the steam separator 24, and the steam dryer 25, so that the reactor dome pressure slightly increases. Further, when the reactor recirculation system 28 is operated in the manual mode, the PLR pump speed and the PLR pump flow rate do not change.

  When the lower shroud portion (mainly the weld line H7 portion) is damaged, a part of the coolant (liquid phase) of the core lower plenum 31 flows directly from the damaged portion of the core shroud 11 to the lower portion of the downcomer portion 32. . For this reason, since the coolant from the core lower plenum 31 to the core 18 decreases, the in-channel flow rate through which the coolant flows in the fuel assembly of the core 18 decreases, and as a result, the thermal neutron flux decreases and APRM Decreases. As the thermal neutron flux decreases, the fission reaction of the fuel assembly decreases, so the amount of steam generated decreases, and the reactor dome pressure in the upper dome 26 of the reactor pressure vessel 17 decreases.

  Further, when a part of the coolant in the lower core plenum 31 flows out, the pressure in the lower core plenum 31 is lowered, and thus the core differential pressure is lowered. However, since the core flow rate is measured as a total value of discharge flow rates of a plurality of jet pumps 30 or a total value of discharge flow rates of a plurality of recirculation internal pumps (RIP) (not shown), it actually decreases. However, it is measured with an apparent increase. Further, when the reactor recirculation system 28 is operated in the manual mode, the PLR pump speed and the PLR pump flow rate do not change.

  In the database unit 13, in addition to the short-term fluctuation pattern of the physical quantity in the boiling water reactor 16 as described above, the long-term fluctuation pattern in units of days, months, or years includes an average value of the physical quantity in a certain period. Are stored as at least one statistic of variation range and standard deviation. Thereby, for example, the fluctuation tendency of the physical quantity when the damage (for example, the through hole) of the core shroud 11 gradually increases from a slight one is stored in the database unit 13.

  Thus, the physical quantity fluctuation pattern in the boiling water reactor 16 stored in the database unit 13 is the accumulation of physical quantities measured by the actual measurement unit 12 in the present embodiment. Therefore, the physical quantity variation pattern when a new breakage actually occurs in the boiling water reactor 16 is stored in the database unit 13, and the database unit 13 has a learning function.

  The comparison / determination unit 14 compares the variation pattern of the physical quantity in the nuclear reactor measured by the actual measurement unit 12, for example, the boiling water reactor 16, with the variation pattern of the corresponding physical quantity stored in the database unit 13, It is determined whether or not these fluctuation patterns match, and the presence or absence of breakage in the boiling water reactor 16 is detected, and further, the breakage site is specified and detected.

  That is, the comparison / determination unit 14 uses a predetermined criterion (for example, the change rate of the physical quantity or the like) based on the variation pattern of the physical quantity measured by the actual measurement unit 12 and the variation pattern of the corresponding physical quantity obtained from the database unit 13. , Change amount, etc.) and if both fluctuation patterns match, it is determined that there is damage in the reactor (for example, the core shroud 11). At the time of this comparison, since the fluctuation pattern of the physical quantity is stored in the database unit 13 for each damaged part, the comparison determination unit 14 determines whether the damaged pattern matches the fluctuation pattern, thereby causing the damage. It becomes possible to identify and grasp the part.

  For example, the axial portion of the core shroud 11 in the axial direction of damage can be basically specified by a variation pattern such as LPRM, APRM, core differential pressure, JP flow rate (measured core flow rate) among the above-described various physical quantities. . Moreover, the radial direction site | part of the fracture | rupture of the core shroud 11 compares LPRM which the neutron flux monitor in the vicinity of a failure site measures, and LPRM which the neutron flux monitor installed in the position facing this neutron flux monitor compares. Is basically identifiable.

  Further, when comparing the above-described variation patterns, the comparison / determination unit 14 quantifies and outputs the degree of variation pattern matching in order to clarify the range of deviation between both variation patterns. For example, the comparison / determination unit 14 digitizes the degree of coincidence between the two variation patterns using a percentage, or directly digitizes and displays the difference between the two variation patterns. The comparison / determination unit 14 outputs information related to breakage, such as the presence / absence of the breakage detected, the broken portion, and the degree of matching described above to the display unit 15.

  The display unit 15 is connected to the comparison / determination unit 14, and displays information related to breakage such as the presence / absence of breakage and a broken part output by the comparison / determination unit 14. For example, the display unit 15 outputs an alarm for notifying that the core shroud 11 in the reactor pressure vessel 17 has been damaged, and displays the damaged part on the display screen.

  Next, the operation will be described with reference to FIG.

  The actual measurement unit 12 measures various physical quantities in the boiling water reactor 16, and the comparison / determination unit 14 acquires a variation pattern of the measured physical quantities from the actual measurement unit 12 (S1).

  The comparison / determination unit 14 outputs the variation pattern of the physical quantity acquired from the actual measurement unit 12 to the database unit 13 and inquires the database unit 13 about a variation pattern that is similar or substantially coincident (S2).

  The database unit 13 outputs a similar or substantially matching variation pattern to the comparison determination unit 14 for the physical quantity corresponding to the physical quantity measured by the actual measurement unit 12 (S3).

  The comparison / determination unit 14 compares the variation pattern of the physical quantity measured by the actual measurement unit 12 with the variation pattern of the corresponding physical quantity output from the database unit 13 using a predetermined criterion (S4). . Then, the comparison / determination unit 14 detects, for example, whether or not the core shroud 11 in the reactor pressure vessel 17 in the boiling water reactor 16 is damaged, identifies the damaged portion, and outputs it to the display unit 15. (S5).

  Therefore, according to the present embodiment, the following effect (1) is obtained.

  (1) The comparison / determination unit 14 is stored in advance in the database unit 13 and the fluctuation pattern of the physical quantity measured during the normal operation by the actual measurement unit 12 in the nuclear reactor 16 during the normal operation of the boiling water reactor 16. The boiling water reactor 16 is compared with, for example, the fluctuation pattern of the physical quantity in the reactor 16 when the core shroud 11 is damaged, and it is determined whether or not these fluctuation patterns are matched. For example, whether or not the core shroud 11 of the nuclear reactor 16 is damaged is determined, and the damaged portion is specified. From this, during normal operation of the boiling water reactor 16, it is possible to quickly detect, for example, the core shroud 11 of the boiling water reactor 16 along with the damaged portion. As a result, it is possible to quickly and accurately determine whether or not the boiling water reactor 16 is stopped or continued in the microsign maintenance. Moreover, in the post-mortem maintenance, since the damaged part is known in advance, quick post-mortem maintenance can be performed.

[B] Second embodiment (FIG. 5)
FIG. 5 is a block diagram showing a second embodiment of the in-reactor damage detection apparatus according to the present invention. In the second embodiment, the same parts as those in the first embodiment are denoted by the same reference numerals, and the description will be simplified or omitted.

  The in-reactor damage detection apparatus 40 of the present embodiment differs from the in-reactor damage detection apparatus 10 of the first embodiment in that the variation pattern of the physical quantity stored in the database unit 13 is a numerical analysis unit. This is a physical quantity data obtained by numerical analysis according to 41.

  In other words, the numerical analysis unit 41 calculates and analyzes the fluctuation patterns of various physical quantities when the core shroud 11 of the boiling water reactor 16 is damaged, for example, by in-core flow analysis or core dynamic characteristic analysis, The analysis result is stored in the database unit 13.

  Since the other points are the same as those of the first embodiment, the present embodiment also has the same effects as the effect (1) of the first embodiment.

[C] Third embodiment (FIG. 6)
FIG. 6 is a block diagram showing a third embodiment of the in-reactor damage detection apparatus according to the present invention. In the third embodiment, the same parts as those in the first embodiment are denoted by the same reference numerals, and the description will be simplified or omitted.

  The in-reactor damage detection apparatus 43 of the present embodiment is different from the in-reactor damage detection apparatus 10 of the first embodiment in that the variation pattern of the physical quantity stored in the database unit 13 is, for example, an experimental measurement value. This is a point that is data of a physical quantity measured by a previously conducted experiment stored in the storage unit 44. The experiment is performed by, for example, a mop-up test.

  Since the other points are the same as those of the first embodiment, the present embodiment also has the same effects as the effect (1) of the first embodiment.

[D] Fourth embodiment (FIGS. 7 and 8)
FIG. 7 is a block diagram showing a fourth embodiment of the in-reactor damage detection apparatus according to the present invention. In the fourth embodiment, the same parts as those in the first embodiment are denoted by the same reference numerals, and the description will be simplified or omitted.

  The in-reactor damage detection device 46 of the present embodiment differs from the in-reactor damage detection device 10 of the first embodiment in that a numerical analysis unit is used instead of the database unit 13 of the in-reactor damage detection device 10. 47.

  The numerical analysis unit 47 obtains, by numerical analysis, a fluctuation pattern of the physical quantity measured by the actual measurement unit 12 when it is assumed that there is no damage in the core shroud 11 of the boiling water reactor 16, for example. In other words, the numerical analysis unit 47 performs an actual measurement by numerically analyzing a phenomenon (for example, in-core flow analysis or core dynamic characteristic analysis) when it is assumed that the core shroud 11 of the boiling water reactor 16 is not damaged. The variation pattern of the physical quantity corresponding to the physical quantity measured by the unit 12 is obtained and output.

  The comparison / determination unit 14 determines in advance the physical quantity variation pattern in the boiling water reactor 16 measured by the actual measurement unit 12 and the corresponding physical quantity variation pattern obtained by the numerical analysis unit 47. Comparison is made using a standard (for example, a change rate or a change amount of a physical quantity), and it is determined whether or not these fluctuation patterns match, thereby detecting the presence or absence of damage in the core shroud 11 of the boiling water reactor 16, for example. Then, a message to that effect is output to the display unit 15. Further, when comparing the above-described variation patterns, the comparison / determination unit 14 quantifies and outputs the degree of variation pattern matching to the display unit 15 in order to clarify the range of deviation between both variation patterns.

  The operation of the present embodiment will be described with reference to FIG.

  The actual measurement unit 12 measures various physical quantities in the boiling water reactor 16, and the comparison / determination unit 14 acquires a variation pattern of the measured physical quantities from the actual measurement unit 12 (S11).

  The comparison / determination unit 14 outputs the variation pattern of the physical quantity measured by the actual measurement unit 12 to the numerical analysis unit 47 and inquires of the numerical analysis unit 47 about the physical quantity (S12).

  The numerical analysis unit 47 obtains and compares the physical quantity corresponding to the physical quantity measured by the actual measurement unit 12 by numerical analysis when it is assumed that no damage has occurred in, for example, the core shroud 11 of the boiling water reactor 16. It outputs to the determination part 14 (S13).

  The comparison / determination unit 14 compares the variation pattern of the physical quantity measured by the actual measurement unit 12 with the variation pattern of the corresponding physical quantity obtained from the numerical analysis unit 42 using a predetermined criterion (S14). ). Then, the comparison / determination unit 14 detects whether or not the boiling water reactor 16 is broken, for example, whether or not the core shroud 11 is broken, and outputs it to the display unit 15 (S15).

  Therefore, according to the present embodiment, the following effect (2) is obtained.

  (2) The comparison / determination unit 14 sets numerical values for the physical quantity variation pattern measured during the normal operation by the actual measurement unit 12 in the nuclear reactor 16 during normal operation of the boiling water reactor 16 and the measured physical quantity. The fluctuation pattern obtained by numerical analysis by the analysis unit 47 is compared with the fluctuation pattern when it is assumed that the core shroud 11 of the boiling water reactor 16 is not damaged, and whether or not these fluctuation patterns match. Judgment is made to determine whether or not, for example, the core shroud 11 of the boiling water reactor 16 is damaged. Thus, during normal operation of the boiling water reactor 16, it is possible to quickly detect, for example, damage to the core shroud 11 of the boiling water reactor 16.

  As mentioned above, although this invention was demonstrated based on the said embodiment, this invention is not limited to this.

  For example, in each embodiment, the case where the core shroud 11 of the boiling water reactor 16 is damaged has been described. However, other parts other than the core shroud 11, such as the shroud head 22, the stand pipe 23, the steam The present invention is similarly applied to the case where the separator 24 or the shroud leg 34 provided in the reactor pressure vessel 17 for supporting the lower end portion of the core shroud 11 on the reactor pressure vessel 17 is damaged. Can be applied.

  Further, in each of the first to third embodiments, the database unit 13 describes a physical quantity variation pattern in the case where the core shroud 11 of the boiling water reactor 16 is damaged, for example. However, similarly to the case of the fourth embodiment, the fluctuation pattern of the physical quantity when the boiling water reactor 16 is not damaged may be stored.

  Furthermore, in each embodiment, although the case of the boiling water reactor 16 was described, this invention is applicable also to the case of a pressurized water reactor.

1 is a block diagram showing a first embodiment of a reactor damage detection apparatus according to the present invention. Sectional drawing which shows the boiling water nuclear reactor for demonstrating the physical quantity measured in the damage detection apparatus in a reactor of FIG. The chart which shows the fluctuation pattern of the main physical quantity stored in the database part of FIG. 1 for every broken site. The flowchart which shows the effect | action in the reactor internal damage detection apparatus of FIG. The block diagram which shows 2nd Embodiment of the damage detection apparatus in a reactor which concerns on this invention. The block diagram which shows 3rd Embodiment of the damage detection apparatus in a reactor which concerns on this invention. The block diagram which shows 4th Embodiment of the damage detection apparatus in a reactor which concerns on this invention. The flowchart which shows the effect | action in the nuclear reactor breakage detection apparatus of FIG.

Explanation of symbols

DESCRIPTION OF SYMBOLS 10 Reactor internal damage detection apparatus 11 Core shroud 12 Measurement part 13 Database part 14 Comparison judgment part 15 Display part 16 Boiling water reactor 17 Reactor pressure vessel 40 Reactor damage detection apparatus 41 Numerical analysis part 43 Reactor internal damage Detection device 44 Experimental measurement value storage unit 46 Reactor damage site detection device 47 Numerical analysis unit

Claims (14)

An in-reactor damage detection device that detects damage in the reactor during normal operation of the reactor,
An actual measurement unit for measuring a physical quantity in the reactor during normal operation of the reactor;
A database unit for preliminarily storing a fluctuation pattern of a physical quantity in the nuclear reactor when the nuclear reactor is damaged or does not exist;
Compare the fluctuation pattern of the physical quantity in the nuclear reactor measured by the actual measurement section with the fluctuation pattern of the corresponding physical quantity stored in the database section, and determine whether or not these fluctuation patterns match. And a comparison / determination unit for determining whether or not there is damage in the nuclear reactor. In the database unit, the fluctuation pattern of the physical quantity in the nuclear reactor when there is damage in the nuclear reactor is stored in advance for each damaged portion,
The comparison / determination unit determines whether or not the variation pattern of the physical quantity measured by the actual measurement unit matches the variation pattern of the corresponding physical quantity stored in the database unit. The in-reactor damage detection apparatus according to claim 1, wherein a damage site is specified together with presence or absence. An in-reactor damage detection device that detects damage in the reactor during normal operation of the reactor,
An actual measurement unit for measuring a physical quantity in the reactor during normal operation of the reactor;
For the physical quantity measured in this actual measurement unit, a numerical analysis unit for obtaining a fluctuation pattern by numerical analysis when it is assumed that no damage exists in the nuclear reactor,
Compare the fluctuation pattern of the physical quantity in the reactor measured by the actual measurement section with the fluctuation pattern of the corresponding physical quantity obtained by the numerical analysis section, and determine whether or not these fluctuation patterns match. And a comparison / determination unit for determining whether or not there is damage in the reactor. The physical quantities in the reactor are the local neutron flux, core average neutron flux, core flow rate, core differential pressure, jet pump flow rate, recirculation pump flow rate, recirculation pump speed, reactor dome pressure, reactor bottom drain temperature. The in-reactor breakage detecting device according to any one of claims 1 to 3, wherein the at least one of the pressure difference between the inside and outside of the shroud and the temperature difference of the coolant inside the reactor is at least one of the following. 5. The reactor internal damage according to claim 1, wherein the fluctuation pattern of the physical quantity stored in the database part is a storage of the physical quantity measured by the actual measurement part. Detection device. The in-reactor damage detection device according to claim 1, wherein the fluctuation pattern of the physical quantity stored in the database unit is data of a physical quantity obtained by numerical analysis. 5. The in-reactor damage detection apparatus according to claim 1, wherein the fluctuation pattern of the physical quantity stored in the database unit is data of a physical quantity measured by a previously conducted experiment. . 8. The physical quantity variation pattern stored in the database unit includes at least one statistic of an average value, a variation range, and a standard deviation of physical quantities in a certain period. The in-reactor damage detection device according to any one of the above. The comparison / determination unit matches the variation pattern of the physical quantity measured by the actual measurement unit with the variation pattern of the corresponding physical quantity stored in the database unit or obtained by the numerical analysis unit. The in-reactor damage detection apparatus according to any one of claims 1 to 8, wherein the damage detection apparatus has a function of digitizing and outputting the degree of. 10. A display unit is connected to the comparison / determination unit, and the display unit is configured to be able to display the presence / absence of breakage, a damaged part, etc. output from the comparison / determination unit. The in-reactor damage detection device according to any one of the above. A method for detecting damage in a reactor that detects damage in the reactor during normal operation of the reactor,
Preliminarily storing the fluctuation pattern of the physical quantity in the reactor when the reactor is damaged or not present in the database unit,
Compare the fluctuation pattern of the physical quantity measured in the reactor during normal operation of the nuclear reactor with the fluctuation pattern of the corresponding physical quantity stored in the database unit, and determine whether these fluctuation patterns match. And determining whether there is damage in the reactor and detecting the damage in the reactor. In the database unit, the fluctuation pattern of the physical quantity in the nuclear reactor when there is damage in the nuclear reactor is stored in advance for each damaged portion,
Judgment is made whether the fluctuation pattern of the physical quantity measured in the nuclear reactor matches the fluctuation pattern of the corresponding physical quantity stored in the database section, and the damaged portion is identified along with the presence or absence of the damage in the nuclear reactor. The in-reactor damage detection method according to claim 11, wherein damage in the reactor is detected. A method for detecting damage in a reactor that detects damage in the reactor during normal operation of the reactor,
Measure physical quantities in the reactor during normal operation of the reactor,
About this measured physical quantity, the fluctuation pattern when assuming that there is no damage in the reactor is obtained by numerical analysis,
Compare the measured physical quantity fluctuation pattern described above with the corresponding physical quantity fluctuation pattern obtained by numerical analysis, determine whether these fluctuation patterns match, and determine whether there is damage in the reactor. And detecting a damage in the nuclear reactor. The physical quantities in the reactor are the local neutron flux, core average neutron flux, core flow rate, core differential pressure, jet pump flow rate, recirculation pump flow rate, recirculation pump speed, reactor dome pressure, reactor bottom drain temperature. The method for detecting damage within a reactor according to any one of claims 11 to 13, wherein the method is at least one of a differential pressure inside and outside the shroud and a temperature difference between coolants in the reactor.

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