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Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles
Nuclear Engineering and Design 144 (1993) 389-397 North-Holland
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Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles * T.L. Chapman, S. R a n g a n a t h a n d G . L . S t e v e n s General Electric Nuclear Energy, 175 Curtner Avenue, t a l c 747, San Jose, CA 95125, USA Received 9 November 1992, revised version 30 April 1993
As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.
1. Introduction Cracking was observed in boiling water reactor (BWR) feedwater nozzles during the 1970s. These cracks were attributed to high cycle thermal fatigue resulting from the mixing of the colder feedwater flow with the higher temperature reactor water on the nozzle surface. The original feedwater sparger designs were potentially subject to cracking initiated by thermal fatigue because of a slip-fit thermal sleeve design. * Extended and updated version of the presentation given at the SMiRT-11 Post-Conference Seminar No. 2, Assuring Structural Integrity of Steel Reactor Pressure Boundary Components, Taipei, Taiwan, August 26-28, 1991. 0029-5493/93/$06.00
The crack initiation was the result of rapid temperature cycling associated with the turbulent mixing of relatively cold feedwater that leaked past the slip fit. The presence of stainless steel cladding on the inner surface also caused higher thermal stresses which aggravated the cracking phenomenon. As a result of extensive evaluation, G E Nuclear Energy ( G E N E ) developed new thermal sleeve designs with reduced leakage and r e c o m m e n d e d that the cladding be removed from the nozzle inner radius and bore, thereby reducing the high-cycle fatigue susceptibility and improving ultrasonic inspectability. In addition, changes in sparger design, specific system modifications, and changes in operational procedures were implemented to further mitigate crack initiation and growth. Operating B W R
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T.L. Chapman et al. /Application o[" advanced ultrasonic test techniques
plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of the acceptability of the design changes, the Nuclear Regulatory Commission (NRC) issued NUREG-0619 [1]. The NRC indicated in NUREG-0619 that the design changes are responsive to the issue and provide an acceptable approach to nozzle crack mitigation. However, to account for unexpected crack initiation or the presence of previous indications, NUREG-0619 also required that a crack growth evaluation be performed, and established guidelines for periodic ultrasonic (UT) and liquid pcnetrant (PT) testing. The purpose of the program described herein is to combine effective, automated ultrasonic examinations with on-line fatigue monitoring systems to provide a sound technical basis for supplanting the PT exams required by NUREG-0619. A further intent of this program is to instigate a procedure for potentially reducing the frequency of UT examinations in the future. The inspection technique implements special methods of ultrasonic illumination of the nozzle inner surfaces to provide data on crack location and size. The methods include: (1) scanning with optimized transducers and techniques from the vessel wall surfaces to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that yield the crack location, size (i.e., arc-length) and maximum depth of penetration into the nozzle inner surface. The method is most accurate for large cracks with a detection probability near unity. For smaller cracks with low penetration, the detection probability is lower, as is the accuracy of the measured crack characteristics. The program objective was to be as deterministic and quantitative as practically achievable with existing field application constraints and considering personnel radiation exposure reduction (ALARA) objectives. These techniques have been developed to the point where they are now considered a reliable alternative to the PT requirements of NUREG-0619. Coupled with the UT inspections, a combined program of automated fatigue, leakage and crack growth monitoring was implemented for the feedwater nozzle region. This approach provides for a continuous assessment of the integrity and the susceptibility to high cycle thermal fatigue crack initiation. This assessment of the nozzle structure is made by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage
and performing fracture mechanics predictions based on actual thermal duty, in the event a crack indication is found by inspection. Collectively, the monitoring provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.
2. Technical approach The intent of the requirements set forth in NUREG-0619 were two-fold: (1) use UT examinations to detect the deeper cracks which have propagated by subsequent temperature and pressure cycling well into the nozzle base metal, and (2) use PT examinations to detect the small and shallow fatigue cracks which have initiated but not been driven to depths detectable by UT. These requirements were based on the examination methods and techniques available and routinely used at the time the N U R E G was written (1980). The objectives of the program were as follows: (i) to improve upon the first (UT) requirement described above, i.e., implement advanced, automated UT techniques to detect and size not only larger flaws, but also the shallow indications that could be due to thermal fatigue alone. (ii) substitute the second (PT) requirement with on-line monitoring to provide assurance concerning crack initiation. By monitoring leakage flow, fatigue usage and fatigue crack growth on a real-time basis, it is possible to verify the absence of shallow thermal fatigue cracks, thereby avoiding the need for PT examination and the associated personnel exposure, Advances in UT examination methods and techniques over the time period since NUREG-0619, coupled with operating experience and advanced, automated monitoring systems can be effectively combined to eliminate many of the contingencies required by NUREG-0619. Such an approach can also significantly eliminate or reduce other concerns. For example, the significant radiation exposure involved with the RPV drain down and PT examination of the nozzle region can be eliminated with such a program. Automated UT examinations which use state-ofthe-art methods and techniques, when properly combined with fatigue monitoring systems, provide a sound technical basis for supplanting feedwater nozzle PT examinations in BWRs and reducing the frequency of future UT examinations. Figures 1 and 2 graphically compare the proposed approach to the present requirements of NUREG-0619. The proposed approach presumes the adequacy of inspection, monitoring and
T.L. Chapman et al. /Application of adcanced ultrasonic test techniques
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Fig. 2. Current NUREG-0619 requirements (triple sleeve sparger design).
Since one of the objectives of the program was to eliminate the need for PT examinations, it was important to assure that mechanisms that could lead to shallow surface cracking were monitored and evaluated. Since thermal fatigue is a key cracking mechanism for the feedwater nozzle, it is important to monitor fatigue usage in the nozzle. The design basis for the feedwater nozzle includes assumptions regarding operational parameters such as pressure, flow, temperature, and various other technical factors. The fatigue usage is based on design basis cycles, which describe the amplitude and frequency of the pressure and temperature transients anticipated for the component lifetime. If it could be assured that the actual plant parameters will never exceed the design basis cycles and that all transient conditions are fully anticipated, the fatigue usage would be known apriori and there would be no need to monitor plant transients. In reality, actual plant transients differ from those assumed in the design basis, both in terms of temperature change magnitudes and frequency of occurrence. For example, GENE has found from the analysis of several plant histories that the actual number of transients (such as startup/shutdown or SCRAM) experienced by operating reactors has been higher than that anticipated in the design basis. However, many of the events experienced by the reactor are significantly less severe than that assumed in the design basis. Since these plant transients contribute to the cumulative fatigue usage of key reactor components, monitoring of the significant transients in order to assess the longterm life and integrity for susceptible vessel components is important. Past experience with feedwater nozzle inner radius cracking has clearly identified this region as one of the more susceptible areas of the BWR vessel. Accurate evaluations of fatigue duty and thermal sleeve seal condition can be obtained using the General Electric Fatigue Monitoring System (GEFMS). The GEFMS has the capability of uninterrupted data acquisition, as well as the ability to perform several monitoring tasks concurrently. As a result, the GEFMS is a system which will perform three important functions to address the feedwater nozzle integrity issue: (1) Fatigue Duty Monitoring, (2) Leakage Monitoring,
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and (3) Crack Growth Monitoring. Each of these functions is briefly described below.
The analysis software used by the fatigue duty module utilizes the influence function approach. This approach is graphically depicted in Figure 4. Since this approach starts with a detailed finite element model of the actual nozzle configuration, the resulting stresses are as accurate as typical design basis stress analyses (see Figure 5). The key benefit is that the system uses actual plant data obtained from plant instrumentation already in place. Therefore, the resulting fatigue usage results provide a more realistic assessment of the actual duty experienced by the component. This directly addresses the issue of the difference between actual plant duty and the duty assumed in the design basis, as described above.
3.1. Fatigue monitoring The fatigue duty portion of the GEFMS is a module which will calculate and keep track of actual fatigue usage for the feedwater nozzle. This software consists of a data acquisition module, which continuously collects measured plant parameters tapped from existing plant instrumentation, and an analysis module which calculates the stresses and fatigue duty for the feedwater nozzle based on the collected data. The overall system is shown in Figure 3. Step Temperature Change AT
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Fig. 4. Influence function methodology.
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T.L. Chapman et aL /Application of advanced ultrasonic test techniques 3.2. Leakage monitoring Leakage of feedwater flow past the thermal sleeve seals is important since it can mix with the hotter reactor water and causes high frequency thermal cycling along the face of the nozzle. Thus, monitoring the bypass leakage flow is a key step in assuring the absence of fatigue cracks. The leakage portion of the GEFMS is a module which will calculate and monitor thermal sleeve bypass leakage for the feedwater nozzles. Knowledge of the bypass leakage is important in that the integrity of the thermal sleeve seals can be inferred from it. With the integrity of the thermal sleeve seals maintained, the high cycle fatigue condition associated with the temperature cycling caused by leakage is low or non-existent, and confidence is high that cracks will not initiate. The leakage monitoring software consists of a data acquisition module, which collects measured nozzle surface temperatures, and an analysis module which calculates leakage flow based on the collected data. This module requires the installation of resistance temperature detectors (RTDs) at key locations on the nozzles. The signals from these RTDs are multiplexed and fed into the GEFMS computer, where they are processed. This module and its interaction with the overall GEFMS is shown in Figure 6. The analysis software used by the leakage module utilizes an analytical relationship between the measured temperature downstream of the secondary seal, (by use of R T D - T 1 in Figure 6) and leakage flow. This analytical relationship is substantiated by a data base accumulated by G E N E during extensive testing of the feedwater nozzles several years ago. A portion of this testing was specifically aimed at the development of a system for accurately measuring secondary seal leakage flow in the triple thermal sleeve sparger design. The temperature is normalized to the feedwater temperature (by use of R T D - T 2 in Figure 6) so that changes in reactor conditions can be accounted for.
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The crack growth portion of the GEFMS is a module which will calculate and track fatigue crack growth of assumed (or suspected) flaws in the feedwater nozzle. The primary purpose of this module is to demonstrate that the stresses at the feedwater nozzle corner, calculated in the fatigue duty module described above, do not result in growth to greater than preestablished allowable levels for the intended period of operation.
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Fig. 6. Thermal sleeve leakage monitoring process.
The crack growth monitoring software uses the measured plant information as well as the stresses already determined by the fatigue duty module. An analysis module is provided which predicts crack growth using this information along with an approach consistent with ASME Section XI methodology. Figure 7 is an example of crack growth analysis for an initial 0.25-inch deep flaw. The results of the crack growth module are key to demonstrating continued operation for an extended period of time in the event that an indication is found during UT examinations.
Fatigue Crack Growth Results Predicted Crack Growth as a Function of Time 1.0 0.9 0.8 0.7 0.6
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4. Ultrasonic inspection technique Early in-service inspections of nuclear plants using UT methods utilized manual scanning and data recording techniques normally performed in high radiation zones where scan time was necessarily kept to a minimum. Nozzle inner surfaces were both clad and unclad, and inspection entailed long metal paths. These conditions, combined with the problems associated with precise positioning, resulted in a quantity and accuracy of data typically less than desired, which reduced the confidence in the examination results. These and other factors prompted the periodic internal PT inspection of feedwater nozzles as specified in NUREG-0619. In the GENE program which developed automated UT data acquisition systems and techniques, emphasis was placed on methods which have significantly improved flaw detection and characterization. The UT techniques were developed and tested on full-size RPV nozzle mockups of various sizes and designs. Notches in these mockups, located in the nozzle inner radius and bore, ranged in depth from 0.114 to 0.750 inch. Figure 8 depicts the various zones of the nozzle. Technique development was based on extensive mockup testing to determine which combination of parameters was best suited for flaw detection (i.e., the ability to resolve small flaw signals from noise signals). These parameters included, but were not limited to, location of the UT beam entry surface, beam angle with respect to the entry surface, ultrasonic beam mode (longitudinal or transverse), pulse-echo or pitch-catch data gen-
eration mode, and the angle at which the beam intersected the flaw. The data analysis software included a 3-dimensional graphics package that superimposed peak UT data points on a nozzle image, such as Figure 9. The software allowed viewing data from various orientations and at various magnifications. Access to digitized Ascan data was available for viewing of any selected data point, which greatly aided in the classification and evaluation of UT data. The inspection method for the nozzle inner radius and zone 2A of Figure 8 is illustrated in Figure 10. By pulsing a transmitter at a specified angle-of-incidence
Fig. 10. Ray interactions with a nozzle flaw - 3-D view.
T.L. Chapman et aL/ Application of advanced ultrasonic test techniques
with respect to the surface, refracted longitudinal-waves are directed towards the zone under inspection. A raster scanning path provides increased probability of interacting with the flaw and returning energy to the receiver. By positioning the receiver at a specified distance and orientation with respect to the transmitter and the examination zone, conditions can be made favorable for receiving a reflected pulse from a flaw in the material volume. For zones 2B and 3 of Figure 8, single element shear-wave transducers may be used in the scanner in a conventional way, or refracted longitudinal waves may be used in the pitch-catch mode. The electronic display of the returned signal, embodied in a digitized A-Scan, is shown in Figure 11. In the A-scan display, the amplitude (height) of the re-
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turned signals is shown on the vertical axis, and the relative time-of-flight is shown on the horizontal axis. Note the high signal-to-noise ratio typified by the data. The signal amplitude is affected by the relative orientation of the sonic beam direction and the flaw aspect, as determined by the transducer position. The locations of the various flaws are indicated on the nozzle planview, and the relative amplitude of the returns are color-coded for ease of interpretation. Extensive mockup testing showed that this nozzle inspection method was more effective than previous techniques, because: - The system design provided improved dynamic range (i.e., higher signal-to-noise ratio responses), and consequent improvements in small flaw detection capability;
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- The system provided for multiple, redundant coverage of all zones in both clockwise and counter-clockwise directions; - Multiple angles of refraction were provided by the raster scan pattern to enhance flaw detection probability; - Enhanced data acquisition, storage and retrieval capabilities, as well as A-, B- and C-scan displays, provided for a larger data base and a more reliable baseline for future inspection comparisons. As a result of these techniques, G E N E demonstrated increased examination accuracy and reliability on full-size nozzle mockups. The integrity of the feedwater nozzles could therefore be confirmed by automatic ultrasonic inspection, defect detection and characterization, and fracture mechanics analysis. This method is now considered to be a reliable alternative to the PT examination requirement of NUREG-0619.
5. U l t r a s o n i c
technique
qualification
Appendix VIII, Supplement 5, "Qualification Requirements for Inside Radius Examinations" provided rules for extending a qualification for examination of the clad-base metal interface on the vessel (Supplement 4) to a nozzle inside radius by using a mockup containing some additional notches. Supplement 5 stated that the specimens shall comply with Supplement 4, except that the flaws may be either cracks or notches. For the case of the nozzle inner radius examination, notches were considered equally representative. Because this qualification plan was concerned with only the unclad nozzle inner radius, the requirements for the size and number of flaws were adopted directly from Supplement 4. The minimum sample set required at least seven flaws for detection qualification, and an additional three flaws to qualify the sizing technique (i.e., a total of ten were required). While Supplement 4 allowed flaws up to 0.750 inch in depth to be used in the sample set, the GENE qualification plan had a maximum flaw depth of 0.550 inch, a more conservative condition than required. Some notches were wider than nominally specified by Appendix VIII, but this was not considered detrimental to the qualification. The influence of notch width was not a significant advantage for the G E N E technique. The notches were not filled; however, this did not to impact the ultrasonic examination. The flaw configuration in the feedwater nozzle mockup placed flaws in all inspection zones for the qualification data set. Flaws were radially oriented as specified in Appendix VIII, which was also
plan
The ultrasonic technique qualification plan was based on testing full-scale mockups. The plan incorporated a data sample set developed using ASME Code, Section XI, Appendix VIII as a guideline. Appendix VIII did not contain specific rules for qualification of inside radius examination methods for unclad nozzles, but it was used as a guide to evaluate data, define sizing methodology and devise field inspection procedures.
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Fig. 12. Feedwater nozzle mockup flaw placement.
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T.L. Chapman et al. /Application of advanced ultrasonic test techniques in accordance with fracture mechanics predictions and field experience with nozzle cracks. Figure 12 shows a typical development of the mockup notch layout. In this case, the majority of notches were in zones 2 and 3. Other mockups contained additional zone 1 notches, as well as actual fatigue crack implants, which were also be used in the qualification. The primary purpose of this demonstration was to qualify the equipment and technique. Development of the full protocol for an Appendix VIII qualification is an industry effort that is still on-going at the present time. Use of existing mock-ups with flaws placed in the various inspection zones of Figures 8 and 12 was considered sufficient, in the absence of protocol qualification. Automatic data recording and retrieval capability allowed for subsequent reviews of inspection data, as required. Personnel were trained in the UT technique using the samples, and those performing data analysis were given a practical examination using recorded data, similar to that used under SNT-TC-LA qualification programs. During the first field examinations, these personnel included the development specialists responsible for the technique and equipment. The qualification plan was focused on specific portions of Appendix VIII that were applicable to the feedwater nozzle inspection. Flaw depths in the range of 0.15 to 0.35 inch were in the sample set, encompassing the 0.25-inch basis in NUREG-0619. The sizing procedure provided by Appendix VIII was applied to the flaw data and generated a measurement which could be compared to the actual notch depth. Since the data was automatically recorded, it was available for subsequent scrutiny and review for adequacy.
6. Conclusions
The GENE approach to feedwater nozzle inner radius and bore flaws was to ultrasonically inspect to locate and quantify any anomalies present, then to utilize automated monitoring systems and analytical methods to determine fatigue and predict crack growth rates. Leakage monitoring was implemented to assure that thermal sleeve-to-seal integrity is maintained within the 0.3 to 0.5 gallons per minute guideline of NUREG-0619. Re-inspections at appropriate intervals are now to be used to confirm that flaw sizes are within predicted and acceptable limits. Since no flaws were detected and leak rates were acceptable, ultrasonic inspections should be repeated at regular intervals consistent with
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the ASME Code requirements. It was recommended and approved that feedwater sparger removal and PT inspection was not necessary with the implementation of these systems. In the unlikely event excessive leak rates and crack growth rates would have been detected and confirmed, seal refurbishment a n d / o r crack repair procedures would have been planned and implemented for the future and an immediate drain-down and PT would have been avoided. The application of modern ultrasonic testing methods to feedwater nozzle inspection is now well within the state-of-the-art, and GENE techniques developed over the past few years provide the means of performing quantitative inspections routinely. Automated monitoring systems, coupled with fracture mechanics analyses, provide the means of assessing crack initiation and growth conditions, as well as the impact of any flaws discovered by inspection. Continued plant operation can be technically justified in the event that nozzle cracks are detected, based upon crack sizes and growth rates determined by quantitative measurements and analysis methods now available. The application of this program was completed for a domestic BWR-4 plant in the Fall of 1991. Final NRC concurrence and approval of the approach was obtained by October 1991, and full endorsement and completion of the program was implemented by December 1991.
References
[1] NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (November 1980). [2] NRC Generic Letter 81-11 to all Power Reactor Licensees from Darrell Eisenhut (February 28, 1981). [3] NED0-21821-01, 79NED252, Class I, Boiling Water Reactor Feedwater Nozzle/Sparger Final Report (Supplement 1), General Electric Company (January 1979). [4] NED0-24327, 81NED257, Class I, General Electric Boiling Water Reactor Feedwater Nozzle Surveillance Instrumentation System, General Electric Company (April 1981). [5] G.L. Stevens and S. Ranganath, Use of On-line Fatigue Monitoring of Nuclear Reactor Components as a Tool for Plant Life Extension, Life Assessment and Life Extension of Power Plant Components - 1989, PVP Volume 171, pp. 85-92, The 1989 ASME Pressure Vessels and Piping Conference, Honolulu, July 23-27, 1989, ASME, New York. [6] ASME Boiler and Pressure Vessel Code, Section XI, American Society of Mechanical Engineers, New York. [7] ASME Boiler and Pressure Vessel Code, Section III, American Society of Mechanical Engineers, New York.
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Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles * T.L. Chapman, S. R a n g a n a t h a n d G . L . S t e v e n s General Electric Nuclear Energy, 175 Curtner Avenue, t a l c 747, San Jose, CA 95125, USA Received 9 November 1992, revised version 30 April 1993
As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.
1. Introduction Cracking was observed in boiling water reactor (BWR) feedwater nozzles during the 1970s. These cracks were attributed to high cycle thermal fatigue resulting from the mixing of the colder feedwater flow with the higher temperature reactor water on the nozzle surface. The original feedwater sparger designs were potentially subject to cracking initiated by thermal fatigue because of a slip-fit thermal sleeve design. * Extended and updated version of the presentation given at the SMiRT-11 Post-Conference Seminar No. 2, Assuring Structural Integrity of Steel Reactor Pressure Boundary Components, Taipei, Taiwan, August 26-28, 1991. 0029-5493/93/$06.00
The crack initiation was the result of rapid temperature cycling associated with the turbulent mixing of relatively cold feedwater that leaked past the slip fit. The presence of stainless steel cladding on the inner surface also caused higher thermal stresses which aggravated the cracking phenomenon. As a result of extensive evaluation, G E Nuclear Energy ( G E N E ) developed new thermal sleeve designs with reduced leakage and r e c o m m e n d e d that the cladding be removed from the nozzle inner radius and bore, thereby reducing the high-cycle fatigue susceptibility and improving ultrasonic inspectability. In addition, changes in sparger design, specific system modifications, and changes in operational procedures were implemented to further mitigate crack initiation and growth. Operating B W R
© 1993 - E l s e v i e r S c i e n c e P u b l i s h e r s B.V. A l l r i g h t s r e s e r v e d
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T.L. Chapman et al. /Application o[" advanced ultrasonic test techniques
plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of the acceptability of the design changes, the Nuclear Regulatory Commission (NRC) issued NUREG-0619 [1]. The NRC indicated in NUREG-0619 that the design changes are responsive to the issue and provide an acceptable approach to nozzle crack mitigation. However, to account for unexpected crack initiation or the presence of previous indications, NUREG-0619 also required that a crack growth evaluation be performed, and established guidelines for periodic ultrasonic (UT) and liquid pcnetrant (PT) testing. The purpose of the program described herein is to combine effective, automated ultrasonic examinations with on-line fatigue monitoring systems to provide a sound technical basis for supplanting the PT exams required by NUREG-0619. A further intent of this program is to instigate a procedure for potentially reducing the frequency of UT examinations in the future. The inspection technique implements special methods of ultrasonic illumination of the nozzle inner surfaces to provide data on crack location and size. The methods include: (1) scanning with optimized transducers and techniques from the vessel wall surfaces to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that yield the crack location, size (i.e., arc-length) and maximum depth of penetration into the nozzle inner surface. The method is most accurate for large cracks with a detection probability near unity. For smaller cracks with low penetration, the detection probability is lower, as is the accuracy of the measured crack characteristics. The program objective was to be as deterministic and quantitative as practically achievable with existing field application constraints and considering personnel radiation exposure reduction (ALARA) objectives. These techniques have been developed to the point where they are now considered a reliable alternative to the PT requirements of NUREG-0619. Coupled with the UT inspections, a combined program of automated fatigue, leakage and crack growth monitoring was implemented for the feedwater nozzle region. This approach provides for a continuous assessment of the integrity and the susceptibility to high cycle thermal fatigue crack initiation. This assessment of the nozzle structure is made by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage
and performing fracture mechanics predictions based on actual thermal duty, in the event a crack indication is found by inspection. Collectively, the monitoring provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.
2. Technical approach The intent of the requirements set forth in NUREG-0619 were two-fold: (1) use UT examinations to detect the deeper cracks which have propagated by subsequent temperature and pressure cycling well into the nozzle base metal, and (2) use PT examinations to detect the small and shallow fatigue cracks which have initiated but not been driven to depths detectable by UT. These requirements were based on the examination methods and techniques available and routinely used at the time the N U R E G was written (1980). The objectives of the program were as follows: (i) to improve upon the first (UT) requirement described above, i.e., implement advanced, automated UT techniques to detect and size not only larger flaws, but also the shallow indications that could be due to thermal fatigue alone. (ii) substitute the second (PT) requirement with on-line monitoring to provide assurance concerning crack initiation. By monitoring leakage flow, fatigue usage and fatigue crack growth on a real-time basis, it is possible to verify the absence of shallow thermal fatigue cracks, thereby avoiding the need for PT examination and the associated personnel exposure, Advances in UT examination methods and techniques over the time period since NUREG-0619, coupled with operating experience and advanced, automated monitoring systems can be effectively combined to eliminate many of the contingencies required by NUREG-0619. Such an approach can also significantly eliminate or reduce other concerns. For example, the significant radiation exposure involved with the RPV drain down and PT examination of the nozzle region can be eliminated with such a program. Automated UT examinations which use state-ofthe-art methods and techniques, when properly combined with fatigue monitoring systems, provide a sound technical basis for supplanting feedwater nozzle PT examinations in BWRs and reducing the frequency of future UT examinations. Figures 1 and 2 graphically compare the proposed approach to the present requirements of NUREG-0619. The proposed approach presumes the adequacy of inspection, monitoring and
T.L. Chapman et al. /Application of adcanced ultrasonic test techniques
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Since one of the objectives of the program was to eliminate the need for PT examinations, it was important to assure that mechanisms that could lead to shallow surface cracking were monitored and evaluated. Since thermal fatigue is a key cracking mechanism for the feedwater nozzle, it is important to monitor fatigue usage in the nozzle. The design basis for the feedwater nozzle includes assumptions regarding operational parameters such as pressure, flow, temperature, and various other technical factors. The fatigue usage is based on design basis cycles, which describe the amplitude and frequency of the pressure and temperature transients anticipated for the component lifetime. If it could be assured that the actual plant parameters will never exceed the design basis cycles and that all transient conditions are fully anticipated, the fatigue usage would be known apriori and there would be no need to monitor plant transients. In reality, actual plant transients differ from those assumed in the design basis, both in terms of temperature change magnitudes and frequency of occurrence. For example, GENE has found from the analysis of several plant histories that the actual number of transients (such as startup/shutdown or SCRAM) experienced by operating reactors has been higher than that anticipated in the design basis. However, many of the events experienced by the reactor are significantly less severe than that assumed in the design basis. Since these plant transients contribute to the cumulative fatigue usage of key reactor components, monitoring of the significant transients in order to assess the longterm life and integrity for susceptible vessel components is important. Past experience with feedwater nozzle inner radius cracking has clearly identified this region as one of the more susceptible areas of the BWR vessel. Accurate evaluations of fatigue duty and thermal sleeve seal condition can be obtained using the General Electric Fatigue Monitoring System (GEFMS). The GEFMS has the capability of uninterrupted data acquisition, as well as the ability to perform several monitoring tasks concurrently. As a result, the GEFMS is a system which will perform three important functions to address the feedwater nozzle integrity issue: (1) Fatigue Duty Monitoring, (2) Leakage Monitoring,
392
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and (3) Crack Growth Monitoring. Each of these functions is briefly described below.
The analysis software used by the fatigue duty module utilizes the influence function approach. This approach is graphically depicted in Figure 4. Since this approach starts with a detailed finite element model of the actual nozzle configuration, the resulting stresses are as accurate as typical design basis stress analyses (see Figure 5). The key benefit is that the system uses actual plant data obtained from plant instrumentation already in place. Therefore, the resulting fatigue usage results provide a more realistic assessment of the actual duty experienced by the component. This directly addresses the issue of the difference between actual plant duty and the duty assumed in the design basis, as described above.
3.1. Fatigue monitoring The fatigue duty portion of the GEFMS is a module which will calculate and keep track of actual fatigue usage for the feedwater nozzle. This software consists of a data acquisition module, which continuously collects measured plant parameters tapped from existing plant instrumentation, and an analysis module which calculates the stresses and fatigue duty for the feedwater nozzle based on the collected data. The overall system is shown in Figure 3. Step Temperature Change AT
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T.L. Chapman et aL /Application of advanced ultrasonic test techniques 3.2. Leakage monitoring Leakage of feedwater flow past the thermal sleeve seals is important since it can mix with the hotter reactor water and causes high frequency thermal cycling along the face of the nozzle. Thus, monitoring the bypass leakage flow is a key step in assuring the absence of fatigue cracks. The leakage portion of the GEFMS is a module which will calculate and monitor thermal sleeve bypass leakage for the feedwater nozzles. Knowledge of the bypass leakage is important in that the integrity of the thermal sleeve seals can be inferred from it. With the integrity of the thermal sleeve seals maintained, the high cycle fatigue condition associated with the temperature cycling caused by leakage is low or non-existent, and confidence is high that cracks will not initiate. The leakage monitoring software consists of a data acquisition module, which collects measured nozzle surface temperatures, and an analysis module which calculates leakage flow based on the collected data. This module requires the installation of resistance temperature detectors (RTDs) at key locations on the nozzles. The signals from these RTDs are multiplexed and fed into the GEFMS computer, where they are processed. This module and its interaction with the overall GEFMS is shown in Figure 6. The analysis software used by the leakage module utilizes an analytical relationship between the measured temperature downstream of the secondary seal, (by use of R T D - T 1 in Figure 6) and leakage flow. This analytical relationship is substantiated by a data base accumulated by G E N E during extensive testing of the feedwater nozzles several years ago. A portion of this testing was specifically aimed at the development of a system for accurately measuring secondary seal leakage flow in the triple thermal sleeve sparger design. The temperature is normalized to the feedwater temperature (by use of R T D - T 2 in Figure 6) so that changes in reactor conditions can be accounted for.
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The crack growth monitoring software uses the measured plant information as well as the stresses already determined by the fatigue duty module. An analysis module is provided which predicts crack growth using this information along with an approach consistent with ASME Section XI methodology. Figure 7 is an example of crack growth analysis for an initial 0.25-inch deep flaw. The results of the crack growth module are key to demonstrating continued operation for an extended period of time in the event that an indication is found during UT examinations.
Fatigue Crack Growth Results Predicted Crack Growth as a Function of Time 1.0 0.9 0.8 0.7 0.6
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4. Ultrasonic inspection technique Early in-service inspections of nuclear plants using UT methods utilized manual scanning and data recording techniques normally performed in high radiation zones where scan time was necessarily kept to a minimum. Nozzle inner surfaces were both clad and unclad, and inspection entailed long metal paths. These conditions, combined with the problems associated with precise positioning, resulted in a quantity and accuracy of data typically less than desired, which reduced the confidence in the examination results. These and other factors prompted the periodic internal PT inspection of feedwater nozzles as specified in NUREG-0619. In the GENE program which developed automated UT data acquisition systems and techniques, emphasis was placed on methods which have significantly improved flaw detection and characterization. The UT techniques were developed and tested on full-size RPV nozzle mockups of various sizes and designs. Notches in these mockups, located in the nozzle inner radius and bore, ranged in depth from 0.114 to 0.750 inch. Figure 8 depicts the various zones of the nozzle. Technique development was based on extensive mockup testing to determine which combination of parameters was best suited for flaw detection (i.e., the ability to resolve small flaw signals from noise signals). These parameters included, but were not limited to, location of the UT beam entry surface, beam angle with respect to the entry surface, ultrasonic beam mode (longitudinal or transverse), pulse-echo or pitch-catch data gen-
eration mode, and the angle at which the beam intersected the flaw. The data analysis software included a 3-dimensional graphics package that superimposed peak UT data points on a nozzle image, such as Figure 9. The software allowed viewing data from various orientations and at various magnifications. Access to digitized Ascan data was available for viewing of any selected data point, which greatly aided in the classification and evaluation of UT data. The inspection method for the nozzle inner radius and zone 2A of Figure 8 is illustrated in Figure 10. By pulsing a transmitter at a specified angle-of-incidence
Fig. 10. Ray interactions with a nozzle flaw - 3-D view.
T.L. Chapman et aL/ Application of advanced ultrasonic test techniques
with respect to the surface, refracted longitudinal-waves are directed towards the zone under inspection. A raster scanning path provides increased probability of interacting with the flaw and returning energy to the receiver. By positioning the receiver at a specified distance and orientation with respect to the transmitter and the examination zone, conditions can be made favorable for receiving a reflected pulse from a flaw in the material volume. For zones 2B and 3 of Figure 8, single element shear-wave transducers may be used in the scanner in a conventional way, or refracted longitudinal waves may be used in the pitch-catch mode. The electronic display of the returned signal, embodied in a digitized A-Scan, is shown in Figure 11. In the A-scan display, the amplitude (height) of the re-
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turned signals is shown on the vertical axis, and the relative time-of-flight is shown on the horizontal axis. Note the high signal-to-noise ratio typified by the data. The signal amplitude is affected by the relative orientation of the sonic beam direction and the flaw aspect, as determined by the transducer position. The locations of the various flaws are indicated on the nozzle planview, and the relative amplitude of the returns are color-coded for ease of interpretation. Extensive mockup testing showed that this nozzle inspection method was more effective than previous techniques, because: - The system design provided improved dynamic range (i.e., higher signal-to-noise ratio responses), and consequent improvements in small flaw detection capability;
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T.L. Chapman et al. /Application of advanced ultrasonic test techniques
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- The system provided for multiple, redundant coverage of all zones in both clockwise and counter-clockwise directions; - Multiple angles of refraction were provided by the raster scan pattern to enhance flaw detection probability; - Enhanced data acquisition, storage and retrieval capabilities, as well as A-, B- and C-scan displays, provided for a larger data base and a more reliable baseline for future inspection comparisons. As a result of these techniques, G E N E demonstrated increased examination accuracy and reliability on full-size nozzle mockups. The integrity of the feedwater nozzles could therefore be confirmed by automatic ultrasonic inspection, defect detection and characterization, and fracture mechanics analysis. This method is now considered to be a reliable alternative to the PT examination requirement of NUREG-0619.
5. U l t r a s o n i c
technique
qualification
Appendix VIII, Supplement 5, "Qualification Requirements for Inside Radius Examinations" provided rules for extending a qualification for examination of the clad-base metal interface on the vessel (Supplement 4) to a nozzle inside radius by using a mockup containing some additional notches. Supplement 5 stated that the specimens shall comply with Supplement 4, except that the flaws may be either cracks or notches. For the case of the nozzle inner radius examination, notches were considered equally representative. Because this qualification plan was concerned with only the unclad nozzle inner radius, the requirements for the size and number of flaws were adopted directly from Supplement 4. The minimum sample set required at least seven flaws for detection qualification, and an additional three flaws to qualify the sizing technique (i.e., a total of ten were required). While Supplement 4 allowed flaws up to 0.750 inch in depth to be used in the sample set, the GENE qualification plan had a maximum flaw depth of 0.550 inch, a more conservative condition than required. Some notches were wider than nominally specified by Appendix VIII, but this was not considered detrimental to the qualification. The influence of notch width was not a significant advantage for the G E N E technique. The notches were not filled; however, this did not to impact the ultrasonic examination. The flaw configuration in the feedwater nozzle mockup placed flaws in all inspection zones for the qualification data set. Flaws were radially oriented as specified in Appendix VIII, which was also
plan
The ultrasonic technique qualification plan was based on testing full-scale mockups. The plan incorporated a data sample set developed using ASME Code, Section XI, Appendix VIII as a guideline. Appendix VIII did not contain specific rules for qualification of inside radius examination methods for unclad nozzles, but it was used as a guide to evaluate data, define sizing methodology and devise field inspection procedures.
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T.L. Chapman et al. /Application of advanced ultrasonic test techniques in accordance with fracture mechanics predictions and field experience with nozzle cracks. Figure 12 shows a typical development of the mockup notch layout. In this case, the majority of notches were in zones 2 and 3. Other mockups contained additional zone 1 notches, as well as actual fatigue crack implants, which were also be used in the qualification. The primary purpose of this demonstration was to qualify the equipment and technique. Development of the full protocol for an Appendix VIII qualification is an industry effort that is still on-going at the present time. Use of existing mock-ups with flaws placed in the various inspection zones of Figures 8 and 12 was considered sufficient, in the absence of protocol qualification. Automatic data recording and retrieval capability allowed for subsequent reviews of inspection data, as required. Personnel were trained in the UT technique using the samples, and those performing data analysis were given a practical examination using recorded data, similar to that used under SNT-TC-LA qualification programs. During the first field examinations, these personnel included the development specialists responsible for the technique and equipment. The qualification plan was focused on specific portions of Appendix VIII that were applicable to the feedwater nozzle inspection. Flaw depths in the range of 0.15 to 0.35 inch were in the sample set, encompassing the 0.25-inch basis in NUREG-0619. The sizing procedure provided by Appendix VIII was applied to the flaw data and generated a measurement which could be compared to the actual notch depth. Since the data was automatically recorded, it was available for subsequent scrutiny and review for adequacy.
6. Conclusions
The GENE approach to feedwater nozzle inner radius and bore flaws was to ultrasonically inspect to locate and quantify any anomalies present, then to utilize automated monitoring systems and analytical methods to determine fatigue and predict crack growth rates. Leakage monitoring was implemented to assure that thermal sleeve-to-seal integrity is maintained within the 0.3 to 0.5 gallons per minute guideline of NUREG-0619. Re-inspections at appropriate intervals are now to be used to confirm that flaw sizes are within predicted and acceptable limits. Since no flaws were detected and leak rates were acceptable, ultrasonic inspections should be repeated at regular intervals consistent with
397
the ASME Code requirements. It was recommended and approved that feedwater sparger removal and PT inspection was not necessary with the implementation of these systems. In the unlikely event excessive leak rates and crack growth rates would have been detected and confirmed, seal refurbishment a n d / o r crack repair procedures would have been planned and implemented for the future and an immediate drain-down and PT would have been avoided. The application of modern ultrasonic testing methods to feedwater nozzle inspection is now well within the state-of-the-art, and GENE techniques developed over the past few years provide the means of performing quantitative inspections routinely. Automated monitoring systems, coupled with fracture mechanics analyses, provide the means of assessing crack initiation and growth conditions, as well as the impact of any flaws discovered by inspection. Continued plant operation can be technically justified in the event that nozzle cracks are detected, based upon crack sizes and growth rates determined by quantitative measurements and analysis methods now available. The application of this program was completed for a domestic BWR-4 plant in the Fall of 1991. Final NRC concurrence and approval of the approach was obtained by October 1991, and full endorsement and completion of the program was implemented by December 1991.
References
[1] NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (November 1980). [2] NRC Generic Letter 81-11 to all Power Reactor Licensees from Darrell Eisenhut (February 28, 1981). [3] NED0-21821-01, 79NED252, Class I, Boiling Water Reactor Feedwater Nozzle/Sparger Final Report (Supplement 1), General Electric Company (January 1979). [4] NED0-24327, 81NED257, Class I, General Electric Boiling Water Reactor Feedwater Nozzle Surveillance Instrumentation System, General Electric Company (April 1981). [5] G.L. Stevens and S. Ranganath, Use of On-line Fatigue Monitoring of Nuclear Reactor Components as a Tool for Plant Life Extension, Life Assessment and Life Extension of Power Plant Components - 1989, PVP Volume 171, pp. 85-92, The 1989 ASME Pressure Vessels and Piping Conference, Honolulu, July 23-27, 1989, ASME, New York. [6] ASME Boiler and Pressure Vessel Code, Section XI, American Society of Mechanical Engineers, New York. [7] ASME Boiler and Pressure Vessel Code, Section III, American Society of Mechanical Engineers, New York.