On-line monitoring and analysis of reactor vessel integrity

• US 4,801,421 A
• Filed: 03/16/1987
• Issued: 01/31/1989
• Est. Priority Date: 06/04/1985
• Status: Expired due to Fees

• Current Assignee: Westinghouse Electric Company LLC (Brookfield Asset Management Incorporated)
• Original Assignee: Westinghouse Electric Company LLC (Brookfield Asset Management Incorporated)
• Inventors: Ackerson, D. Scott, Impink, Albert J. Jr., Balkey, Kenneth R., Andreychek, Timothy S.
• Primary Examiner(s): Kyle, Deborah L.
• Assistant Examiner(s): WASIL, DANIEL D
• Application Number:
  US07/026,501
• Time in Patent Office: 687 Days
• Field of Search: 376/247, 376/249, 376/254, 376/255, 376/259, 364/527
• US Class Current: 376/249
• CPC Class Codes: G21C 13/08 Vessels characterised by the material; Selection of materials for pressure vessels
• CPC Class Codes: G21C 17/00 Monitoring; Testing measuring in general G01; Maintaining
• CPC Class Codes: G21C 17/108 Measuring reactor flux
• CPC Class Codes: Y02E 30/30 Nuclear fission reactors

Abstract

A real time analysis of nuclear reactor vessel integrity is preformed by monitoring reactor coolant temperature and pressure and fast neutron fluence, and generating therefrom a visual comparison of the reference nil-ductility transition temperature, RT_ndt, required for flaw initiation and the actual RT_ndt through the entire depth of the vessel wall at selected critical locations. When flaw initiation is indicated at any depth, the RT_ndt required for flaw propagation arrest is also generated and displayed. A display can also be generated to provide guidance for optimum performance during heatup and cooldown. Thermocouples located at the reactor core inlet are used in conjunction with the resistance temperature detectors, RTDs, conventionally provided in the cold leg conduits to generate a more accurate measurement of the coolant temperature at the critical locations in the vessel. Under stagnant flow conditions a correlation is applied to the RTD readings to provide the best estimate of coolant temperature at the critical locations.

Claims

1. A method of real-time on-line monitoring and analysis of nuclear reactor pressure vessel integrity in a unit in which reactor coolant is circulated along the inner wall of the pressure vessel, said method comprising the steps of:

generating signals representative of the temperature of the reactor coolant at selected critical points along said inner wall of the pressure vessel even under stagnant flow conditions on a real-time basis;

generating signals representative of the pressure of the reactor coolant on a real-time on-line basis; and

generating on a real-time on-line basis, as a function of changes in the temperature and pressure signals, a visual representation of a margin to non-ductile failure of the vessel at selected critical points for all flaw depths through the vessel wall.

2. The method of claim 1 wherein said step of generating temperature signals comprises measuring the reactor coolant temperature at a point spaced from the selected critical points along the inner wall of the pressure vessel, storing a correlation for converting the measured temperature to actual coolant temperature at said selected critical points under stagnant flow conditions of the reactor coolant, generating the temperature signals for the selected critical points under stagnant flow conditions by correlating the temperature at those points to the measured temperature using the stored correlation, and generating the temperature signals from the temperature measurements without the correlation otherwise.

3. The method of claim 2 including the steps of measuring the reactor coolant temperature at a first location before the reactor coolant is circulated along the inner wall of the pressure vessel and at a second location after the reactor coolant has circulated along the inner wall of the pressure vessel, comparing the temperature measurements at said two locations and using said correlation with at least one of said temperature measurements to generate said temperature signals when the two measurements differ by a preselected amount indicative of stagnant flow conditions.

4. The method of claim 3 wherein the reactor includes a reactor core inside the pressure vessel with a core inlet at the bottom thereof and a cold leg conduit through which reactor coolant enters the pressure vessel near the top of the reactor core, circulates down along the inner wall of the pressure vessel and passes through the core inlet into the core, and wherein said first temperature measurement is taken in said cold leg conduit and the second temperature measurement is taken at the core inlet.

5. The method of claim 4 wherein said temperature signals are generated from the average between the first and second temperature measurements when said measurements differ by less than said preselected amount.

6. The method of claim 4 wherein said reactor has a plurality of reactor coolant loops each including a cold leg conduit and wherein said method includes taking a first temperature measurement for each loop in the associated cold leg conduit and taking a second temperature measurement for each loop at the core inlet under the associated cold leg conduit and wherein said method further includes the step of preselecting the temperature measurements for generating the temperature signal for a selected critical point from the loop which has the greatest affect on the temperature at the selected point.

7. The method of claim 2 wherein the point spaced from a selected critical point along the inner wall of the pressure vessel at which the reactor coolant temperature is measured is in the cold leg through which reactor coolant enters the pressure vessel and circulates down along said inner wall, wherein said stored correlation is a correlation between a normalized measured temperature and a normalized temperature at said selected critical point, and wherein said correlating step includes determining the normalized measured temperature, determining the normalized temperature at the selected critical point from the normalized measured temperature using said stored correlation, and determining the temperature at the selected critical point from the normalized temperature at such point.

8. The method of claim 7 wherein safety injection water is injected into the cold leg at a known temperature, wherein said step of normalizing the measured reactor coolant temperature comprises determining the ratio of the difference between the measured reactor coolant temperature in the cold leg after safety injection and the temperature of the safety injection water to the difference between the measured reactor coolant temperature in the cold leg before safety injection and the temperature of the safety injection water, and wherein the step of determining the actual temperature at the selected critical point comprises determining the ratio between the sum of the normalized temperature at the selected critical point and the temperature of the safety injection water to the difference between the measured reactor coolant temperature before injection and the temperature of the safety injection water.

9. Apparatus for on-line monitoring and real-time analysis of nuclear reactor pressure vessel integrity in a reactor unit in which reactor coolant is circulated along the inner wall of the pressure vessel, said apparatus comprising:

means for measuring on a real-time on-line basis the temperature of the reactor coolant along the inner wall of the pressure vessel from measurements taken at a point spaced from the inner wall of the vessel even under stagnant flow conditions;

means for measuring on a real-time on-line basis the pressure of the reactor coolant;

digital computer means for periodically sampling said reactor coolant temperature and pressure measurements and for generating on a real-time on-line basis, as a function of changes over time in the measured temperature and pressure, vessel condition signals representative of the proximity to non ductile failure of the pressure vessel wall at selected critical locations for all flaw depths through the pressure vessel wall;

a visual display device; and

means for applying said vessel condition signals generated by the digital computer to the visual display device to generate a real-time on-line visual representation of the proximity of the pressure vessel wall to non-ductile failure at the critical locations for all flaw depths.

10. The apparatus of claim 9 in combination with:

a reactor core in the reactor pressure vessel having a core inlet at the bottom thereof; and

cold leg conduit means connected to the pressure vessel near the top of reactor core for introducing into the pressure vessel reactor coolant which circulates down along the inner wall of the pressure vessel and then upward through the core inlet into the reactor core; and

wherein said means for measuring reactor coolant temperature includes first sensor means for measuring the temperature of the reactor coolant in the cold leg conduit means and second sensor means for measuring reactor coolant temperature in said core inlet and wherein said digital computer means periodically samples the temperature measurements from both said first and second sensor means and when the temperature measurements from the first and second sensor means differ by more than a predetermined amount indicative of stagnant reactor coolant flow, applies a stored correlation ot the measurements from at least one of said sensor means to generate the temperature measurement of the reactor coolant at the inner wall of the pressure vessel for use in generating said vessel condition signals.

11. The apparatus of claim 10 wherein said cold leg conduit means includes a plurality of cold leg conduits angularly distributed around the reactor pressure vessel, each forming a part of a reactor coolant loop and each having first sensor means therein for measuring the reactor coolant temperature, and wherein said digital computer means compares the reactor coolant temperature measurement from the first sensor means in each cold leg conduit with the temperature measurements from second sensor means substantially vertically aligned therewith to determine a reactor coolant loop temperature for each loop and utilizes the loop temperature which most influences each critical location in the pressure vessel wall for generating the pressure vessel condition signal therefor.

12. The apparatus of claim 11 in combination with guide tubes which extend upward through the bottom of the reactor pressure vessel, through the core inlet and into the reactor core, and wherein said second sensor means in said core inlet are located in said guide tubes.

13. The apparatus of claim 12 wherein said second sensor means are thermocouples mounted in a plurality of guide tubes distributed across the reactor core in a pattern such that at least two thermocouples provide redundant core inlet temperature measurements for each reactor coolant loop.

14. A method of on-line monitoring and analysis of nuclear reactor pressure vessel integrity in a unit in which reactor coolant is circulated along the inner wall of the pressure vessel, said method comprising the steps of:

generating on an on-line basis, temperature signals representative of the temperature of the reactor coolant circulating along the inner wall of the pressure vessel;

generating on an on-line basis, a pressure signal representative of the reactor coolant pressure;

generating a signal representative of fast neutron fluence to which the reactor pressure vessel has been subjected;

generating as a function of said fluence signal a visual representation of the actual real time reference nil-ductility transition temperature (RT_ndt) across the entire pressure vessel wall thickness at a preselected critical location in said wall;

generating as a function of transients in said reactor coolant temperature and pressure signals, a visual representation of the real time required RT_ndt, across the entire pressure vessel wall thickness at said selected critical location, said required RT_ndt being the RT_ndt that would be required in the pressure vessel wall for flaw initiation to occur as a result of stresses set-up by said transients; and

superimposing said visual representations of the real-time actual and required RT_ndt'"'"'s for flaw initiation across the entire pressure vessel wall thickness for the selected critical location to generate a visual representation of the difference in value between the actual and required RT_ndt presented as an RT_ndt margin.

15. The method of claim 14 including the steps of monitoring the RT_ndt margin and when the required RT_ndt at any depth through the vessel wall falls below the actual RT_ndt at that depth, indicating that a flaw at that depth would propagate, generating on a real time basis as a function of the transients in the reactor coolant temperature and pressure signals, a visual representation of the RT_ndt required for arrest of the propagation of the flaw, and superimposing the arrest RT_ndt on the visual representation of the actual RT_ndt and flaw initiation RT_ndt to present a visual comparison of the actual RT_ndt and the arrest RT_ndt.

16. The method of claim 14 including the step of operating the reactor to maintain the RT_ndt required for flaw initiation above the actual RT_ndt across the full thickness of a the pressure vessel wall.

17. The method of claim 14 wherein said step of generating the actual RT_ndt representation comprises the steps of:

storing signals representative of selected physical characteristics of said pressure vessel at said critical location; and

generating the actual RT_ndt representation as a function of said stored physical characteristic signals and said fast neutron fluence signal.

18. The method of claim 17 wherein the step of generating the required RT_ndt representation comprises the steps of:

generating as a function of the reactor coolant temperature and pressure signals, the stored selected physical characteristic signals, and said fast neutron fluence signal, a signal representative of the real-time stress intensity factor through the pressure vessel wall and a signal representative of the real-time flaw initiation toughness through the pressure vessel wall at the selected critical location; and

generating the required RT_ndt representation through the pressure vessel wall by determining the RT_ndt s required to make the stress intensity factor signal equal to the flaw initiation toughness signal for all depths through the pressure vessel wall.

19. The method of claim 18 wherein the generation of the real-time stress intensity signal comprises the steps of:

generating from said reactor coolant temperature and pressure signals, signals representative of real-time temperature and pressure distribution through the pressure vessel wall at said critical location;

generating as a function of said temperature and pressure distribution signals, a signal representative of the distribution of the thermal and pressure stress through the pressure vessel wall;

sequentially generating signals representative of flaws of incrementally increasing depth across the pressure vessel wall at said critical location; and

generating said stress intensity factor signal as a function of said thermal and pressure stress distribution signal and said sequentially generated flaw depth signals.

20. The method of claim 19 including additional steps for monitoring and analyzing pressure vessel integrity during reactor heatup and cooldown, said additional steps including the steps of:

monitoring the reactor coolant temperature and pressure signals; and

when changes in said temperature and pressure signals are within preselected limits for heatup and cooldown;

generating heatup/cooldown real-time temperature and pressure distribution signals through the pressure vessel at said critical location, using said reactor coolant temperature and pressure signals and, with regard to said pressure signals, a first preselected safety factor signal;

generating heatup/cooldown thermal and pressure stress distribution signals as a function of said heatup/cooldown temperature and pressure distribution signals;

generating a heatup/cooldown stress intensity signal as a function of said heatup/cooldown thermal and pressure stress distribution signals for an assumed flaw of preselected dimensions extending outward from the inner surface of said pressure vessel wall at said critical location for cooldown and inward from the outer surface for heatup;

generating a heatup/cooldown flaw initiation toughness signal from said stored pressure vessel physical characteristic signals, and said fast neutron fluence signal taking into account a second safety factor;

generating a heatup/cooldown required RT_ndt signal by determining the RT_ndt which makes said heatup/cooldown stress intensity factor signal for said assumed flaw equal to the heatup/cooldown flaw initiation toughness signal at the depth of the assumed flaw; and

generating a real-time representation of a comparison between said heatup/cooldown required RT_ndt and actual RT_ndt signals.

21. The method of claim 20 including the step of operating the nuclear reactor to maintain the required RT_ndt substantially equal to the actual RT_ndt.

22. A method of on-line monitoring and analysis of the integrity of a vessel in which the inner wall thereof is contacted with a fluid which is subject to varying temperature and pressure conditions, the method comprising the steps of;

generating on an on-line basis signals representative of the temperature of said fluid adjacent the inner wall of the vessel;

generating on an on-line basis signals representative of the pressure of said fluid;

generating signals representative of the actual reference nil-ductility transition temperature (RT_ndt) across the entire vessel wall thickness at a preselected critical location in said vessel wall;

generating as a function of transients in said fluid temperature and pressure signals, signals representative of the real time required RT_ndt for flaw initiation across the entire vessel wall thickness at said selected critical location, said required RT_ndt for flaw initiation being the RT_ndt that would be required in the vessel wall for flaw initiation to occur as a result of stresses set up by said transients; and

comparing said actual RT_ndt signals with the required RT_ndt singals for flaw initiation to generate a representation of the difference in value between the actual and required RT_ndt signals.

23. The method of claim 22 wherein the step of comparing said actual and required RT_ndt signals comprises generating real-time visual representations of the values of said actual and required RT_ndt signals across the vessel wall and presenting both visual RT_ndt representations on one display to generate a visual representation of an RT_ndt margin as the difference between the values of the RT_ndt signals displayed.

24. The method of claim 22 wherein the step of generating the required RT_ndt signal for flaw initiation comprises the steps of:

generating as a function of the on-line fluid temperature and pressure signals, a signal representative of the real-time stress intensity factor through the vessel wall, and a signal representative of the real-time flaw initiation toughness through the vessel wall at the selected critical location; and

generating the required RT_ndt signal for flaw initiation by determining the RT_ndts required to make the real- time stress intensity factor signal equal to the real-time flaw initiation toughness signal for all depths through the vessel wall at said selected critical location.

25. The method of claim 24 wherein the step of generating the real-time stress intensity factor signal comprises the steps of:

generating from said on-line fluid temperature and pressure signals, signals representative of real time temperature and pressure distribution through the vessel wall at said selected critical location;

generating as a function of said temperature and pressure distribution signals, a signal representative of the distribution of the thermal and pressure stress through the vessel wall;

sequentially generating signals representative of flaws of incrementally increasing depth across the vessel wall at said selected critical location; and

generating said real-time stress intensity factor signal as a function of said thermal and pressure stress distribution signal and said sequentially generated flaw depth signals.

26. The method of claim 24 including the additional steps of monitoring the RT_ndt margin and when the required RT_ndt signal for flaw initiation falls below the actual RT_ndt signal for any flaw depth, generating on a real-time basis as a function of the on-line fluid temperature and pressure singals, a signal representative of the required RT_ndt for flaw propagation arrest, and displaying this signal with the actual RT_ndt signal.

27. In combination:

a nuclear reactor having a reactor core housed in a pressure vessel with reactor coolant circulating along the inner surface of the pressure vessel walls and through the reactor core;

means for on-line measurement of the reactor coolant temperature and for generating temperature signals therefrom;

means for on-line measurement of the reactor coolant pressure and for generating a pressure signal therefrom;

digital computer means for periodically sampling said reactor coolant temperature and pressure signals and for generating as a function of said on-line temperature and pressure signals, signals representative for selected locations in the vessel of real-time reference nil-ductility transition temperatures (RT_ndt) required for flaw initiation across the vessel wall as a result of stresses induced by changes in said temperature and pressure signals and for generating signals representative of the actual RT_ndt across the vessel wall at each selected location; and

display means for generating a visual display comparing the real-time RT_ndt signals required for flaw initiation with the actual RT_ndt signals for each selected location.

28. The combination of claim 27 including cold leg conduit means through which the reactor coolant which circulates along the inner surface of the pressure vessel walls enters the pressure vessel and core inlet means at the bottom of the reactor core through which the coolant which cirulates along the vessel walls enters the reactor core and wherein said temperature measurement means includes temperature sensors in said cold leg conduit means and in said core inlet means for generating said temperature signals, and wherein said digital computer means generates said required RT_ndt signals for flaw initiation using signals generated by the temperature sensor means in both the cold leg conduit means and the core inlet means.

29. Apparatus for on-line monitoring and analysis of the integrity of a vessel in which the inner wall thereof is contacted by a fluid which is subjected to varying temperature and pressure conditions, said apparatus comprising:

means for measuring on an on-line basis the temperature of the fluid contacting the vessel wall;

means for measuring on an on-line basis the pressure of the fluid; and

digital computer means for periodically sampling the fluid temperature and pressure measurements and for generating on an on-line basis as a function of said on-line temperature and pressure measurements, signals representative, for at least one selected location along the inner wall of said vessel, of real-time reference nil-ductility transition temperatures (RT_ndt) required for flaw initiation at all depths across the vessel wall as a result of stresses induced by changes in said temperature and pressure measurements, and for generating signals representative of the actual RT_ndt across the vessel wall at said selected location; and

means for comparing the actual RT_ndt signal with the required RT_ndt signal for flaw initiation to generate a representation of the difference in value between the actual and required RT_ndt signals.