ML13317B190
| ML13317B190 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/17/1991 |
| From: | Rosenblum R Southern California Edison Co |
| To: | NRC/IRM |
| References | |
| NUDOCS 9105280074 | |
| Download: ML13317B190 (40) | |
Text
Southern California Edison Company 23 PARKER STREET IRVINE, CALIFORNIA 92718 R.
M. ROSENBLUM May 17, 1991 TELEPHONE MANAGER OF (714) 454-4505 NUCLEAR REGULATORY AFFAIRS U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Steam Generator Inspections San Onofre Nuclear Generating Station, Unit 1 INTRODUCTION On April 21, 1991, after operating for 24 days following the 1990-91 Thermal Shield Support Replacement and Cycle 11 Refueling Outage, San Onofre Unit 1 was shut down to locate and repair steam generator tube leakage. The total leak rate just prior to shut down was about 150 gpd, well below Technical Specification limits for steam generator leakage.
To locate the leakage the secondary sides of all three steam generators were pressurized to about 650 psig and 17 leaking tubes were identified. The leaking tubes and surrounding tubes were tested using eddy current testing to determine the cause of the leakage.
The initial eddy current testing characterized the identified leaks as attributable to leaking sleeves. All 17 leaking tubes were plugged. The eddy current inspection also identified one pluggable indication in each steam generator. Given the small sample of tubes initially inspected and the similarity of the indications (all were between the top of the sleeves and the first tube support), Edison elected to expand the inspection. Over several days our inspection was expanded to include:
- 1) 140 tubes full length - no pluggable indications were identified outside the region between the top of the sleeves and the first tube support.
- 2) 860 unsleeved tubes - one pluggable indication in the roll/roll transition was observed.
- 3) 100% of the sleeved tubes - 102 tubes were identified with ECT indications of >30% through wall, all in the region between the top of the sleeve and the first tube support.
Following identification of the defects, a motorized rotating pancake coil ECT probe was utilized to further characterize the indications. Although all the indications were similar in axial and circumferential extent (0.1 to O.43inches axial and 0.1 to 0.61 inches circumferential) ECT analysis was able to distinguish two types of indications in the data, termed Type A and Type B.
91O5280074 910517 PDR ADOCK 05000206 PDR
Document Control Desk May 17, 1991 TYPE A INDICATIONS A total of 17 Type A indications were identified. Growth studies and other analyses performed by SCE could not exclude the possibility that the Type A defects may exhibit a rapid growth rate. However, our analyses show that if left unplugged, they would result in only minor leakage (approximately 0.271 gpm) and will not lead to tube rupture under design basis conditions. All Type A indications have been plugged.
TYPE B INDICATIONS A total of 85 Type B indications (>30% through-wall degradation) were identified. These indications appear to be typical of minor wastage and growth studies have confirmed minimal progression of these defects. Due to their small size and minimal growth rate these defects are not considered to be significant.
Other Anomalies Identified In addition to the 17 leaking sleeves (which includes one Type A indication),
and 16 tubes plugged due to Type A indications, 7 tubes were plugged in this inspection due to other anomalies, including tube obstruction sufficient to preclude satisfactory ECT inspection, a distorted indication in the roll transition zone, and the presence of two sleeves in a single tube.
CONCLUSION A total of 40 tubes were removed from service in this inspection. As further discussed in the attached report, Edison has established that the defects identified in this inspection are not a safety concern and accordingly SONGS 1 is being returned to power.
If you require any additional information, please contact me.
Sincerely Enclosure cc:
George Kalman, NRC Project Manager, San Onofre Unit 1 J. B. Martin, Regional Administrator, NRC Region V C. W. Caldwell, NRC Senior Resident Inspector, San Onofre Units 1, 2&3 C. D. Townsend, NRC Resident Inspector, San Onofre Unit 1
1991 STEAM GENERATOR INSPECTION RESULTS SAN ONOFRE UNIT 1 DOCKET NO. 50-206
- MAY 1991 SOUTHERN CALIFORNIA EDISON COMPANY ROSEMEAD, CALIFORNIA
e__
Table of Contents Section Paqe I. Introduction 1
II. Inspection Program 2
A. Steam Generator Leak Test 2
B. Eddy Current Test Program 3
III.
Flaw Characterization 5
A. Sample Selection 5
B. Motorized Rotating Pancake Coil (MRPC) Examination 5
C. Type A and Type B Flaws 5
D. Type A Flaw Assessment 5
IV.
Flaw Growth Rate 7
A. Methodology 7
B. Results 7
V. Safety Significance of Flaws 9
A. Type "B" Flaws 9
B. Type "A" Flaws 9
C. Enhanced Primary-to-Secondary Leak Rate Monitoring 10 D. Conclusions 12 VI.
Summary and Conclusions 13 A. Summary of Results 13 B. Conclusions 14 VII.
References 15 Appendix A - Tubesheet Maps Of Tubes Inspected Appendix B - Summary - Motorized Rotating Pancake Coil Eddy Current Test Results Appendix C - Summary Of Tubes Plugged And Reason For Plugging Appendix D - Examples - Motorized Rotating Pancake Coil Eddy Current Test Results 91-U1SG.w51
I.
INTRODUCTION On April 21, 1991, after operating for 24 days following the 1990-91 Thermal Shield Support Replacement and Cycle 11 Refueling Outage, San Onofre Unit 1 was shut down to locate and repair steam generator tube leakage. The total leak rate just prior to shut down was about 150 gpd, well below Technical Specification limits for steam generator-leakage.
To locate the leak(s) the secondary sides of all three steam generators were pressurized to about 650 psig and leaking tubes were identified.
Leaking tubes and surrounding tubes were tested using eddy current testing to determine the cause of the leakage.
The purpose of this report is to document Edison's assessment of the indications discovered during this inspection. Detailed root cause analysis, including the cause of newly discovered flaws and the adequacy of previous steam generator inspections will be the subject of separate evaluations.
Section II of this report describes the leak test and the eddy current test program, including the expanded test programs.Section III describes the characterization of flaws detected during the eddy current test program.Section IV discusses the evaluation of current and previous eddy current data to determine the growth rate of the flaws detected in the eddy current test program.Section V describes the evaluation of the safety significance of the flaws.Section VI summarizes the overall conclusions reached as a result of the tests and evaluation. Finally,Section VII provides a listing of the references cited in this report.
II. INSPECTION PROGRI 9
A. Steam Generator Leak Test
- 1. Testing and Corrective Action The secondary side of all three steam generators was pressurized to approximately 650 pounds per square inch and the primary side of the tubesheet in each steam generator channel head was scanned for leaks using a pan and tilt camera. No leakage was observed in any cold leg channel head. A total of seventeen leaking tubes were identified in the hot leg channel heads. All leaking tubes were sleeved on the hot leg end. The breakdown of leaking tubes per steam generator is shown below:
TUBE LEAK RATE SLEEVE TYPE SG (Row-Col)
(Drops/Minute)
(Upper Joint)
A 36-45 30.0 Mechanical 35-70 3.0 Mechanical 35-68 1.5 Mechanical 9-44 0.2 Brazed 31-61
<0.1 Mechanical 17-17
<0.1 Mechanical B
35-55 60.0 Mechanical 31-31
<0.1 Mechanical C
35-56 8.0 Mechanical 37-41 5.0 Mechanical 34-61 1.1 Mechanical 33-45 0.8 Mechanical 34-46 0.4 Mechanical 35-49
<0.1 Mechanical 32-61
<0.1 Mechanical 12-78
<0.1 Braze Converted 31-63
<0.1 Mechanical All leaking tubes were inspected using eddy current testing over their full length, including the sleeved portion. There was only one eddy current indication which may correlate to leakage.
Steam Generator (SG) "A" tube 36-45 had an 88% through-wall outside diameter indication in the tubing at an axial location of 4.4 inches above the top of the sleeve. All seventeen leaking tubes were removed from service by mechanical plugging.
- 2. Conclusions It is inferred that the observed leakage is associated with the sleeve joints based on the lack of correlatable eddy current indications in all but one tube. The leakage for each leak limiting sleeve was within the design basis as discussed in Reference 1 (210 drops per minute).
B. Eddy Current est Program
- 1. Description The conventional bobbin coil probe was used to provide the best possible assessment of the general condition of the inspected length of the non-sleeved portion of the steam generator tubes.
The magnetically biased bobbin probe was employed to assess the condition of the sleeves inspected. The motorized rotating pancake coil (MRPC) probe was used to supplement the bobbin probe when necessary.
The initial eddy current testing program to investigate leakage included all leaking tubes and tubes adjacent to them. Each tube was tested over its full length, including the sleeved portion.
This program consisted of 51 tubes in SG "A", 18 tubes in SG "B",
and 71 tubes in SG "C".
This testing identified two significant eddy current defects in tubing above sleeve ends. A leaking tube in SG "A", 36-45, had an 88% through-wall outside diameter defect at 4.4 inches above the sleeve end. A tube in SG "C",.35-55, had an 86% through-wall outside diameter defect at 0.75 inches above the sleeve end.
- 2. Program Expansion The eddy current testing program was expanded in response to defects in tubing above sleeve ends. The expansion included all sleeved tubes in all steam generators from the top of the sleeve to the first hot leg support plate. A total of 7107 tubes (71.3%
of the tubes in service) were inspected by the conclusion of eddy current testing. This included approximately 860 unsleeved tubes. The unsleeved tubes were inspected from the hot leg tube end through the first hot leg support plate to examine the corresponding area to that in which defects were found in sleeved tubes.
Appendix A consists of tubesheet maps showing tubes inspected.
Appendix B consists of a summary of significant eddy current results from motorized rotating pancake coil (MRPC) probe testing, of tubes with >30% through-wall indications from the bobbin coil examination. All tubes with >30% through-wall indications were inspected using an MRPC except tube 36-45 in SG "A" and tube 35-55 in SG "C".
These tubes were plugged prior to program expansion to include MRPC testing.
- 3. Results and Corrective Action As a result of leak testing and eddy current testing, a total of 40 tubes were removed from service. This consists of 21 tubes in SG "A", 6 tubes in SG "B", and 13 tubes in SG "C".
Appendix C provides the following detail concerning these tubes:
- A listing of each of these tubes and the reason for plugging.
- An elevation view of the steam generator illustrating the tubing axial location designations used in the listing
- Tubest maps illustrating the locati of each of these tubes within the tube bundle.
Five tubes were plugged because of an inside diameter restriction in the vicinity of the sleeve. It is important to note that none of these tubes had been inspected in the vicinity of the sleeve since the sleeving baseline inspection in 1981.
It is postulated that with the exception of one tube restricted by a foreign object, these restrictions may have been present but not identified during the sleeving baseline inspection.
The restrictions in each of the three tubes in SG "A" were viewed with a videoprobe. Tube 7-60 had a foreign object (a washer) lodged approximately 2.5 inches above the sleeve inlet. Tube 4 24 had an indentation approximately 5 inches above the top of the sleeve, in the parent tubing. The indentation was a nearly perfect triangular conical section, with the circular base covering less than ninety degrees of the circumference of the tube. There is no external structure at that axial location to which denting could be attributed. A review of eddy current data for adjacent tubes did not indicate the presence of a foreign object.. Tube 13-44 had a build up of material, which appeared to be similar to braze slag, in the vicinity of the upper sleeve joint.
The restrictions in the sleeves in SG "B" and SG "C" were not viewed with a videoprobe. However, eddy current data indicates the restrictions in each of these two tubes is in the vicinity of the braze within the upper sleeve joint.
Tube 22-33 in SG "A" was plugged because eddy current data indicated that two sleeves were present (one above the other).
Rather than analyze this tube for operation with a second, unrolled sleeve, this tube was plugged.
Tube 7-3 in SG "A" was plugged due to a distorted bobbin coil eddy current indication and MRPC confirmation of an inside diameter indication in the roll/roll transition region of the tube.
A detailed discussion of eddy current flaw results which prompted plugging of other tubes is provided in Section III of this report.
JII.
FLAW CHARACTERI&JN A. Sample Selection Based on the bobbin coil eddy current test results all but two tubes (these tubes had already been plugged) with flaws >30% in the area of interest (flaws in the tube close to the top of the sleeve) were selected for further evaluation (characterization). This sample included 100 tubes, out of 6261 sleeved tubes examined in all 3 steam generators.
B. Motorized Rotating Pancake Coil (MRPC) Examination Each tube in the 100 tube sample was examined using an MRPC probe.
The MRPC probes used were specially fabricated to allow access through the restriction of a sleeve to the tube above the sleeve.
Each tube was examined over the full axial extent of the flaw. Examples of the results of this examination are provided in Appendix D. A summary of the results of the MRPC examinations are provided in Appendix B. Of the 100 indications evaluated using MRPC data, 8 were identified as having non-quantifiable signals (<50% through-wall), 51 were determined to be in the 30-39% through-wall range, 29 were identified to be in the 40-49% through-wall range and 12 were determined to be
>50% through-wall.
Not all of the flaws identified using bobbin coil ECT were detectable using the MRPC; 55 tubes had no detectable degradation (were NDD) using the MRPC. Among those flaws that were detectable using the MRPC, 21 had a high enough signal-to-noise ratio to allow sizing in the circumferential and axial direction. Where obtained these circumferential and axial dimensions are listed in Appendix B.
C. Type A and Type B Flaws During the course of the evaluation of the MRPC data it became apparent that there were two distinct flaw types. The first type of flaw, which is referred to as the "Type A" flaw, was the least common.
Type A flaws are typically small in area 0.1"-0.43" in axial length by 0.1"-0.61" (15-95 degrees) in circumferential extent. The Type A flaws have "sharp" features. The MRPC graphics in Appendix D for tubes 31-30 and 39-43 in the "A" steam generator are typical of Type A flaws. The majority of the flaws were typically shallower but affected a larger area of the tube. These flaws, referred to as "Type B" flaws, have characteristics that are typical of the phosphate wastage (or thinning) that has been previously experienced at San Onofre and other units. The MRPC graphics for tube 20-48 in "A" steam generator are typical of the Type B flaws. There were two tubes with indications that exhibited some characteristics of each of these types and were designated as.Type AB. For conservatism, these two flaws are being considered as Type A flaws in the remainder of this report.
D. Type A Flaw Assessment Due to the relatively unique characteristics of the Type A flaws and their apparent rapid growth, Westinghouse was requested to evaluate these flaws to predict their cause. Westinghouse was given the MRPC data to evaluate in conjunction with materials, chemistry and thermal hydraulic conditions in the steam generators. The results of this evaluation, aw ough not conclusive, confirme at the most likely cause of the Type A flaws is corrosion due to the local concentration of impurities in the sludge pile on the secondary side of the tubes.
Further, based on the relativel low bobbin coil voltage and high MRPC voltage compared to typical crack signals, it is concluded that the Type A flaws are not cracks.
IV.
FLAW GROWTH RA A. Methodology Prior to.identifying that the flaws were divided into two distinct types, a growth study was initiated for all flaws (>30% through-wall) in the region of interest. Eddy current test results from 1980 through 1990 were reviewed for indications in the region from the top of the sleeves to the first tube support. All of the 1980 data, taken with MIZ-12 equipment on magnetic tapes was reanalyzed to determine if there were any flaws within one inch of-the location of the 1991 flaws. Based on this reanalysis of the 1980 data (the last inspection of all of the tubes of interest) an attempt was made to determine the growth rate of all the flaws >30% through-wall in the affected region of the tubes.
B. Results The preliminary results of the growth study indicated that there were two distinct populations of flaws - those with growth rates in the range of 2-10% per cycle and those with growth rates >10% per cycle.
This result supported the eddy current analysts' conclusion (from MRPC results) that there were two distinct flaw types. However, it was also concluded that, due to the significant difference in the method of collecting the eddy current data between 1980 and 1991, a meaningful growth study could not be done for all of the tubes.
For the 17 tubes with Type A flaws a more detailed assessment of historical eddy current data was done. For each inspection from 1980 to 1991, all inspections of the affected tubes were reviewed and all data that possibly included the region of interest was reanalyzed.
The results are provided in Table IV-1. Although there is limited inspection data available, the presence of signals, even though they are non-quantifiable, at the appropriate axial location in tubes 31 30, 36-45, and 26-70 in SG 'A" indicates that these flaws were present but not detected in the 1990 inspection. This result was confirmed directly comparing eddy current test signals from the 1990 and 1991 inspections. The lack of growth during the 1990-91 outage supports a conclusion that the flaws progressed during operation and may have required three cycles to develop.
It is noted that the eddy current inspections subsequent to 1980 exhibited a reduced sensitivity to the region above the sleeves to the first tube support. Based on the current inspection, future inspections will be enhanced to provide greater sensitivity in this region.
TABLE IV-1 EDDY CURRENT TEST RESULTS (%TW) REANALYZED Steam TUBE Location December March Generator Row-Col (In. ATS) 1991 1990 1988 1988 1985 1984 1982 1980
- 1.
"A" 31-30 10.33 82 NQS NI NIR NIR NI NI NDD
- 2.
19-35 10.38 80 NI NI NI NIR NIR NI 18
- 3.
39-43 9.59 64 NI NI NI NIR NI NI NDD
- 4.
36-45 14.22 88 NQS NI NI NI NI NI NDD
- 5.
17-48 15.87 92 NI N(I NIR NI NI NI NDD
- 6.
38-59 8.21 61 NI NI NI NI NI NI 27
- 7.
35-60 12.23 68 NI NDD NI NI NI NI NDD
- 8.
23-62 10.24 75 NI NI NI NI NI NI NDD
- 9.
28-62 8.62 51 NI NI NI NI NDD NI NDD
- 10.
33-65 9.37 37 NDD NI NI NI NI NI 45
- 11.
26-70 11.31 78 NQS NI NI NIR 25 NI 21
- 1. "B" 18-50 15.20 71 NI NI NI NI NI NI 26
- 2.
36-55 10.78 89 NI NI NI NIR NI NI NDD
- 3.
20-56 13.16 45 NI NI NI NDD NI NI 7
- 1. "C" 20-53 10.70 75 NI NI NI NI NI NIR NDD
- 2.
35-55 13.25 86 NI NI NIR NIR NI NIR NDD
- 3.
21-58 j 9.00 66 NI NI NIR NIR NI NIR 15
.Legend:
NI Not Inspected NIR - Not Inspected in The Region of Interest NQS - Non-Quantifiable Signal NDD -
No Detectable Degradation V. SAFETY SIGNIFICANCE OF FLAWS A. Type B Flaws Type B flaws, as indicated in Section III, are typical of phosphate wastage (or thinning) and are progressing relatively slowly.
Accordingly, these flaws are enveloped by those considered in the current safety analysis for the unit.
B. Type A Flaws Type A flaws have been evaluated to determine the leak rate expected when the flaw first leaks and to determine the effect of a nearly through-wall defect on accident analyses.
- 1. Operating Leak Rate The analytical*methodology used to calculate the leak rate is described below:
The crack opening area is calculated using the method proposed by Tada and Paris (Reference 3).
This area is then used in the leak rate computation. These methods have been previously accepted by the NRC (Reference 4).
A maximum leak rate of.271 GPM is calculated based on normal operating conditions and:
RCS temperature
= 600aF RCS pressure
= 2085 Psi Steam Generator shell pressure = 695 Psi Longitudinal crack
= 0.43 inches This flow rate is less than the Technical Specification limit of 0.3 GPM.
- 2. Accident Analyses A crack stability evaluation was performed for two bounding cracks, a through-wall circumferential crack 0.61" long, and a longitudinal crack 0.43" long. These crack sizes were chosen because they represent the maximum dimensions of the indications found during this tube inspection and are larger than any postulated crack that could be found within indications of this size. The analysis was based on an elastic-plastic fracture mechanics approach with tube stresses calculated based on a maximum accident internal tube pressure of 2500 Psi, and no external pressure.
9 -
To pe m the stability calculations following tube characteristics were used:
Tube material is Inconel 600 Tube diameter, 2r = 0.75" Tube thickness, t = 0.055" Temperature, T = 600'F Young's modulus, E = 25200 Ksi Yield strength, Sy = 27.9 Ksi Ultimate strength, Su = 80 Ksi Flow stress, Sf = 53.95 Ksi Ramberg-Osgood stress-strain constants Alpha = 11.56 n = 2.88 Allowable J-integral Jc= 0.30 Kip-inch/inch squared Allowable maximum J-integral = 5 Kip-inch/inch squared The material properties are minimum values for Inconel 600 at 600OF based on the ASME Code. The stress-strain constants, and the fracture toughness (Reference 5) are based on Type 304 stainless steel.
These properties are considered lower bound values for the Inconel 600 tube material.
The Structural Integrity Associates PcCrack computer program (based on EPRI published data) was used to calculate the applied J-integral. The calculated J values are: 0.0587 for the circumferential crack case, and 0.298 for the longitudinal crack case. A stable non-growing crack is defined when Ja plied < JIc*
These results indicate that, under design basis accident condition, both cracks are stable.
C. Enhanced Primary-to-Secondary Leak Rate Monitoring In response to Reference 2 an enhanced primary-to-secondary leak rate monitoring system was implemented at San Onofre Unit 1. Further, although it was subsequently demonstrated that the Unit was not susceptible to rapidly propagating fatigue cracks, the enhanced primary-to-secondary leak rate monitoring program remains in effect.
The San Onofre Unit 1 enhanced monitoring program meets the criteria for being able to monitor a tube failure with leakage characteristics similar to that which occurred at North Anna and take appropriate actions prior to rupture of the tube. The following provides details of the San Onofre Unit 1 enhanced monitoring program.
- 1. Monitoring Leak Rate Data During Power Operations During full power operation and with primary-to-secondary leak rates less than 15 gpd in any one steam generator, the San Onofre Unit 1 leak rate monitoring program consists of continuous monitoring utilizing the air ejector monitor (R-1215) and the steam generator blowdown monitor (R-1216). The basis of the setpoint for R-1215 is the detection of a 15 gpd primary-to secondary leak rate. Condenser air ejector.and feedwater (tritium) samples are collected twice weekly and leak rates are calculated based on the isotopic analysis of these samples.
10
0 0
With R-1215 inoperable, grab samples are collected at the condenser air ejector once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and leak rates are calculated. With R-1216 inoperable, grab samples are collected from the steam. generator blowdown once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The blowdown samples are analyzed isotopically and the results reviewed for indications of primary to secondary leakage.
- 2. Actions In Response To An Increasing Leak Rate
- a. Upon detection and confirmation of a leak rate greater than 15 gpd in any one steam generator, the following activities are performed daily:
(1) Monitor readings are trended and reviewed.
(2) Primary-to-secondary leak rates are calculated based on R-1215 monitor readings.
(3) Primary-to-secondary leak rates are calculated based on condenser air ejector grab sample isotopic analysis.
Under increasing leak rate conditions, the basis of the setpoint for R-1215 is a primary-to-secondary leak rate increase of 15 gpd.
- b. Upon detection and confirmation of a leak rate greater than 40 gpd in any one steam generator or a leak rate increase of more than 15 gpd in any one steam generator in a single day, the following are performed:
(1) Monitor readings are trended and reviewed once per shift (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).
(2) Primary-to-secondary leak rates are calculated based on R-1215 monitor readings once per shift (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).
(3) Primary-to-secondary leak rates are calculated based on condenser air ejector grab sample isotopic analysis at a minimum of once per day.
With leak rates greater than 40 gpd in any one steam generator and R-1215 inoperable, grab samples are collected at the condenser air ejector once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and leak rates are calculated. With R-1216 inoperable, grab samples are collected from the steam generator blowdown once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The blowdown samples are analyzed isotopically and the results reviewed for indications of primary-to-secondary leakage.
- 3. Actions For A Large Leak Rate If the leak rate approaches 100 gpd in any one steam generator, an evaluation of the need to reduce power or commence shutdown is performed. This evaluation considers the leak rate data to date, the rate of change of the leak rate, and the leak rate measurement uncertainty. This evaluation provides adequate time to reduce power or shutdown should a tube fatigue failure be in 11 -
progress. Implementation of this program is more conservative than would be required by the leak rate-time curve generated from the North Anna event.
D.
Conclusions For both the operating leak rate and accident analyses crack stability evaluations, the results are considered to be conservative and bounding because the flaw (crack) size chosen is equivalent to the largest dimension of the indications found by the eddy current inspection. Since the actual cracks would be expected only at or near the bottom of the indications, any actual cracks would be significantly smaller than those considered in these evaluations.
Further, leak monitoring practices will provide assurance that leakage will be detected at an appropriate time to support safe operation during normal and accident conditions.
12 -
VI.
SUMMARY
AND CONCLUSIONS A. Summary of Results As a result of leakage detected during operation, on April 21, 1991, San Onofre Unit 1 was shut down for inspection. The inspection included a leak test in which 17 sleeved tubes were identified as leaking (6 in "A", 2 in "B" and 9 in "C" steam generator). In addition, as a result of eddy current testing of the leaking tubes and those tubes adjacent to the leaking tubes, several tubes were identified as having flaws above, but within 10 inches of the top of the sleeve. As a result of the discovery of these flaws, 100% of the sleeved tubes in all three steam generators were inspected using bobbin coil eddy current testing in the region of interest, between the top of the sleeve and the first tube support. In addition, about 860 unsleeved tubes (23.3% of the unsleeved tubes in service) were inspected between the hot leg tube end and the first hot leg tube support to determine if the unsleeved tubes were affected by the same degradation mechanism. It was concluded that they were not affected.
Subsequent to the leak location and bobbin coil eddy current testing done to locate leaks and tubing flaws, motorized rotating pancake coil (MRPC) eddy current testing was done to characterize the flaws. As a result, two different tubing flaw types, Type A and Type B, were identified. Evaluation of previous inspection data compared to current data provided only a qualitative assessment of flaw growth, but did confirm that there were two distinct flaw types in the region of interest. An assessment of both Type A and Type B flaws was done to determine their safety significance. It was concluded that although Type A flaws may progress rapidly, their physical characteristics are such that they will leak rather than rupture and that the leakage is well within the limits of current safety analyses.
Regarding Type B flaws, it was concluded that they are directly enveloped by those flaws considered in the current safety analyses for the unit.
Based on the testing and evaluation tubes were removed from service as follows:
Steam Generator "A"
"B" "c"
Leaking Sleeves 5
2 9
Type A and AB Flaws 10 3
3 Leaking Sleeve and Type A Flaw 1
0 0
Restricted Tubes 3
1 1
Double Sleeve 1
0 0
Roll/Roll Transition Cracking 1
0 0
Total 21 6
13 13 -
B. Conclusions The-information provided in Section II through V of this report establishes a basis for concluding that the San Onofre Unit 1 Steam Generators can be operated within current operating/inspection limits.
Accordingly, no further action is required and power operation can be safely resumed.
14 -
VII.
REFERENCES
- 1. "Technical Evaluation Report for Hybrid Sleeve," Westinghouse Electric Corporation Report No. NS-MFSE-81-054 dated March 1981 (Proprietary Version), Submitted by Letter K. P. Baskin (SCE) to D. M. Crutchfield (NRC), "Steam Generator Repair Program," dated March 5, 1981
- 2. NRC Bulletin 88-02, "Rapidly Propagating Fatigue Cracks In Steam Generator Tubes"
- 3. H. Tada and P. C. Paris, "Estimation of Stress Intensity and the Crack Opening Area of a Circumferential and Longitudinal Through-Crack in a Pipe"
- 4. EPRI Report NP5531, "Evaluation of High Energy Pipe Rupture Experiments," January 1988.
15 -
APPENDIX A TUBESHEET MAPS OF TUBES INSPECTED
05/91, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT 1 STEAM GENERATOR: A DATE: 05/12/91 TIME:
11: 16: 25 LOCATION: ALL CRITERIA: TUBES EVALUATED COMPLETELY PER REQUIREMENTS PLUGGED 446 EVAL GUMPLETELY 2907 EL-mAmmmm "40 mm m m mme m s m e m o
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.mmm 100 90 80 70 60 5-
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-1 T-
05/91, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT 1 STEAM GENERATOR: B DATE: 05/12/91 TIME: 09:13:29 LOCATION: ALL CRITERIA: TUBES EVALUATED COMPLETELY PER REQUIREMENTS PLUGGED 466 EVAL COMPLETELY 2127 40 mmm mmm AM a
70
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05/91, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT 1 STEAM GENERATOR: C DATE: 05/12/91 TIME: 09 21:48 LOCATION: ALL CRITERIA: TUBES EVALUATED COMPLETELY PER REQUIREMENTS PLUGGED 505 EVAL COMPLETELY 2073 40 i
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APPENDIX B
SUMMARY
MOTORIZED ROTATING PANCAKE COIL EDDY CURRENT TEST RESULTS
Page 1 of 4 Motorized Rotating Pancake Coil Eddy Current Test Results San Onofre Nuclear Generating Station, Unit 1 (May, 1991)
Location 1991 Data Flaw Size ROW COL (inches Bobbin MRPC Axial Circ Flaw ATS)
(%TW)
(%TW)
Length Length Type Steam Generator A 21 17 7.39 40 0
N/M N/M B
16 21 10.22 42 18 N/M N/M B
18 26 9.82 31 47 N/M N/M B
33 27 5.77 39 NDD n/a n/a B
26 29 5.22 33 NDD n/a n/a B
28 29 11.73.
45 NDD n/a n/a B
31 30 10.33 82 76 0.32 0.37 A
25 31 15.37 40 NDD n/a n/a B
19 35 10.38 80 82 0.23 0.50 A
30 36 14.07 32 12 N/M N/M B
27 41 10.55 NQS NDD n/a n/a B
39 43 9.59 46
.64 0.10 0.56 A
39 47 6.96 NQS NDD n/a n/a B
17 48 15.87 92 53 0.22 0.28 A
20 48 9.24 48 0
N/M N/M B
9 49 9.74 NQS NDD n/a n/a B
17 49 9.35 46 31 N/M N/M B
19 52 13.31 30 NDD n/a n/a B
20 52 15.83 49 NDD N/M N/M B
9 53 9.27 38 NDD n/a n/a B_
19 53 12.70 30 17 N/M N/M
'B 20 53 10.09 48 0
N/M N/M B
21 53 9.44 49 NDD n/a n/a B
23 53 17.84 NQS NDD n/a n/a B
19 54 12.94 44 16 0.21 0.35 B
20 54 13.48 37 16 N/M N/M B
19 55 13.09 34 29 N/M N/M B
34 55 10.76 NQS NDD n/a n/a B
35 55 10.68 46 NDD n/a n/a B
42 55 1.26 47 NDD n/a n/a B
18 58 9.57 41 NDD n/a n/a B
38 59 8.21 61 70 0.43 0.18 A
18 60 10.45 44 NDD n/a n/a B
24 60 8.78 41 0
N/M N/M B
34 60 11.94 35 NDD n/a n/a B
35 60 12.23 68 76 0.21 0.44 A
18 61 11.32 30 25 N/M N/M B
22 62 10.59 39 34 N/M N/M B
23 62 10.24 75 78 0.24 0.33 AB 28 62 8.62 51 46 0.20 0.19 AB 34 62 10.73 46 NDD n/a n/a B
21 63 8.61 36 NDD n/a n/a B
Legend:
N/M -
Not Measurable NQS -
Non Quantifiable Signal n/a -
Not Applicable NDD -
No Detectable Degradation
Page 2 of 4 Motorized Rotating Pancake Coil Eddy Current Test Results San Onofre Nuclear Generating Station, Unit 1 (May, 1991)
Location 1991 Data Flaw Size ROW COL (inches Bobbin MRPC Axial Circ Flaw ATS)
(%TW)
(%TW)
Length Length Type Steam Generator A 24 63 12.88 36 NDD n/a n/a B
26 63 8.63 41 29 0.20 0.31 B
27 63 9.44 32 3
N/M N/M B
19 64 9.06 NQS NDD n/a n/a B
34 64 10.06 32 NDD n/a n/a B
37 64 5.43 42 NDD n/a n/a B
20 65 10.44 37 NDD n/a n/a B
31 65 9.96 44 47 0.33 0.18 B
33 65 9.37 37 15 0.18 0.61 A
17 66 9.50 49 NDD n/a n/a B
26 66 11.85 39 26 N/M N/M B
27 66 10.12 NQS NDD n/a n/a B
32 66 8.53 34
- NDD, n/a n/a B
17 67 11.69 37 NDD n/a n/a B
21 68 12.61 37 40 0.20 0.11 B
29 68 9.90 38 NDD n/a n/a B
21 69 10.54 36 NDD n/a n/a B
26 69 11.35 35 NDD n/a n/a B
20 70 11.63 34 NDD n/a n/a B
26 70 11.31 78 78 0.18 0.60 A
23 71 13.13 43 41 N/M N/M B
18 73 12.56 30 38 N/M N/M B
20 73 8.45 42 0
0.23 0.15 B
28 73 10.09 30 17 N/M N/M B
16 76 11.85 36 NDD n/a n/a B
21 77 9.58 36 NDD n/a n/a B
22 77 11.33 33 NDD n/a n/a B
23 77 9.59 37 NDD n/a n/a B
13 78 9.74 36 NDD n/a n/a B
28 78 6.48 40 14 N/M N/M B
15 81 8.99 NQS 0
N/M N/M B
20 81 9.96 33 NDD n/a n/a B
21 81 9.19 43 NDD n/a n/a B
24 81 6.12 31 35 N/M N/M B
15 82 9.45 37 11 N/M N/M B
Legend:
N/M -
Not Measurable NQS -
Non Quantifiable Signal n/a -
Not Applicable NDD -
No Detectable Degradation
Page 3 of 4 Motorized Rotating Pancake Coil Eddy Current Test Results San Onofre Nuclear Generating Station, Unit 1 (May, 1991)
Location 1991 Data Flaw Size ROW COL (inches Bobbin MRPC Axial Circ Flaw ATS)
(%TW)
(%TW)
Length Length Type Steam Generator B 18 50 15.20 71 75 0.21 0.59 A
36 55 10.78 89 84 0.20 0.36 A
20 56 13.16 45 35 0.08 0.42 A
Legend:
N/M -
Not Measurable NQS -
Non Quantifiable Signal n/a -
Not Applicable.
NDD -
No Detectable Degradation
Page 4 of 4 Motorized Rotating Pancake Coil Eddy Current Test Results San Onofre Nuclear Generating Station, Unit 1 (May, 1991)
Location 1991 Data Flaw Size ROW COL (inches Bobbin MRPC Axial Circ Flaw ATS)
(%TW)
(%TW)
Length Length Type Steam Generator C 13 42 10.24 30 NDD n/a n/a B
18 44 8.78 33 NDD n/a n/a B
20 46 9.92 34 NDD n/a n/a B
19 49 14.60 39 11 N/M N/M B
18 50 13.20 35 NDD n/a n/a B
33 51 13.80 34 NDD n/a n/a B
19 53 10.40 46 NDD n/a n/a B
20 53 10.70 75 68 0.41 0.25 A
21 55 10.40 31 33 0.40 0.28 B
21 58 9.00 66 56 0.36 0.45 A
25 58 10.80 34 NDD n/a n/a B
18 59 8.20 38 NDD n/a n/a B
27 62 10.40 38 NDD n/a n/a B
25 63 7.30 48 NDD n/a n/a B
27 63 7.60 32 NDD n/a n/a B
22 67 8.10 44 NDD n/a n/a B
30 69 7.90 37 24 N/M N/M B
24 71 10.10 31 NDD n/a n/a B
28 72 8.30 38 NDD n/a n/a B
26 73 8.30 36 NDD n/a n/a B
Legend:
N/M -
Not Measurable NQS -
Non Quantifiable Signal n/a -
Not Applicable NDD -
No Detectable Degradation
APPENDIX C
SUMMARY
OF TUBES PLUGGED AND REASON FOR PLUGGING
LISTING OF TUBES PLUGGED AND THE REASON Flaw Size Other Tube Number
% Through-Wall Axial Plugging SG Row - Column or Restriction Location Reason A
7 - 3 DRI/IDI (1)
TEH + 1.8 4 -
24 RESTRICTION(2)
TSH + 12.0 17 -
17 LEAKAGE 31 - 30 82 TSH + 10.3 22 - 33 TWO SLEEVES NOT APPLICABLE 19 - 35 80 TSH + 10.4 39 - 43 46 TSH + 9.6 9 -
44 LEAKAGE 13 - 44 RESTRICTION TSH + 4.0 36 - 45 88 TSH + 8.7 LEAKAGE 17 - 48 92 TSH + 15.9 38 - 59 61 TSH + 8.2 7 - 60 FOREIGN OBJECT TEH + 2.5 35 - 60 68 TSH + 12.2 31 -
61 LEAKAGE 23 -
62 75 TSH + 10.2 28 - 62 51 TSH + 8.6 33 -
65 37 TSH + 9.4 35 -
68 LEAKAGE 26 -
70 78 TSH + 11.3 35 -
70 LEAKAGE B
15 - 29 RESTRICTION TSH + 10.0 31 -
31 LEAKAGE 18 - 50 71 TSH + 15.2 35 -
55 LEAKAGE 36 - 55 89 TSH + 7.8 20 - 56 45 TSH + 13.2 C
5 -
19 RESTRICTION TSH + 4.0 37 -
41 LEAKAGE 33 -
45 LEAKAGE 34 -
46 LEAKAGE 35 -
49 LEAKAGE 20 - 53 75 TSH + 10.7 35 - 55 86 TSH + 7.5 35 -
56 LEAKAGE 21 - 58 66 TSH + 9.2 32 -
61 LEAKAGE 34 -
61 LEAKAGE 31 -
63 LEAKAGE 12 -
78 LEAKAGE Notes:
(1) DRI/IDI - Distorted roll indication per bobbin probe, Inside diameter indication per MRPC probe.
(2) Restricts passage of a 0.460" diameter probe.
(3) Percentage through-wall eddy current shown is the indication based on the differential bobbin probe test.
20 -
SERIES 27 TUBE SUPPORT DIAGRAM
.AV2 AV3 ANTI-VIBRATION BARS TUBE SUPPORT NUMBER 04H 04C 4515
.03H 03C 4 5.25 02H 02C 4
45.25 01H 01C 4s.25
23' TRC 2 o0 0 BRH---- -.--
BRC TEH T CUBE SHEET TEHC HOT LEG COLD LEG TSH -TUBE SHEET HOT TRH-TOP.OF ROLL HOT BRH -BOTTOM OF. ROLL HOT TEH -TUBE END HOT
05/91, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: A DATE: 05/16/91 TIME: 09: 19: 29 LOCATION: TEH+
1.5 TO TEC CRITERIA: TUBES TO BE PLUGGED OBS W/.460" PROBE: 3 TYPE "A" IND. ABOVE SLEEVE: 10 LEAKING SLEEVE/PTP: 5 DISTORTED ROLL IND.: 1 LEAKING SLEEVE W/TYPE "A" IND.: 1 DOUBLE SLEEVE/PTP: 1 40 30 20 10 100 90 80 70 60 50 40 30 20 10 CONAM NUCLEAR.
INC.
05/91, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: B DATE: 05/16/91 TIME:
14: 33: 31 LOCATION: TEH+
1.5 TO TEC CRITERIA: TUBES TO BE PLUGGED PIueeED 4W OBS W/.460" PROBE: 1 TYPE "A" IND. ABOVE SLEEVE: 3 LEAKING SLEEVE/PTP: 2 40 30 20 10 100 90 so 70 60 50 40 30 20 10 j
CONAM NUCLEAR, INC.
05/91, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: C DATE: 05/16/91 TIME:
11: 09: 48 LOCATION: TEH+
1.5 TO TEC CRITERIA: TUBES TO BE PLUGGED PG 1
OBS W/.460" PROBE: 1 TYPE "A" IND. ABOVE SLEEVE: 3 LEAKING SLEEVE/PTP: 9 40 30 20 10 100 90 80 70 60 50 40 30 20 10 CONAM NUCLEAR.
INC.
APPENDIX D EXAMPLES MOTORIZED ROTATING PANCAKE COIL EDDY CURRENT TEST RESULTS
THU 12:57 MAY-04-91 SG 10 ROW 31 COL 30 Lark 20: 1 BP 0177 1: 340 DIFF span 20 rot 244 Vert
- of Scans a:
.31 X 8 Y Scale =
1.0 SVT-Trig Offeet w 25 Filter: Off Points/Scan a 70
_-0.40
.40 360
-0.40 180 0
0 0.40 X Rotation
- 60.0.
X Translation =
0 23.i5v 5id 75%
2 Rotation
- 330.0 Y Translation =
0 Opticl Dis Dk 1A Tlal OpticalDiBC$#=
=Dk' 1A -Tape Cal #=tape0l3A.cal0l
THU 12:57 MAY-04-91 SG 10 ROW 31 COL 30 Lark 20: 1 BP DIrF 1: 340 DIFT span 20 rot 244 Vert a of Scans a 31 X B Y Scale =
1.0 Trig Offset a 25 Tilter: Off SVT Points/Scan a 70 Crack Width =
0.45 in Crack Length=
0.36 in
-0.40..
.4.............
00.40 180-0 0 4 X Rotation u 0.0 X Translation =
0 23.15, Sid 75X Z Rotation 0.0 Y Translation =
0 Optical Disc # = Dk 1A Tape Cal # = tapeO13A.cal01
MON 2:39 MAY-08-91 ID 10 ROW 39 COL 43 Lmrk 1: 340 DIFF 1: 340 DIFF span 11 rot 237 Vert a of Scans
=
57 X B Y Scale =
1.0 Trig Offset =
31 Filter: Orf Points/Scan =
60 SVT--0.62
.6................
36
-0.62 18
- 0.
o i.
T 0.62 X Rotation =
60.0 X Translation =
0 5.80v 66d 64%
2 Rotation =
330.0 Y Translation =
0
+M Optical Disc # = Dk 3
Tape Cal # = tape035A.cal01
MON 2:39 MAY-08-91 ID 10 ROW 39 COL 43 Lmrk 1: 340 DIFF 1: 340 DIFF span 11 rot 237 Vert I of Scans =
57 X a Y Scale =
1.0 Trig Offset =
31 Filter: Off Points/Scan =
60 Crack Width =
0.56 in Crack Length=
0.10 in SVT--0.62.
3.62.
-0.62 0
0.62 180 0
X Rotation =
0.0 X Translation =
0 5.+0v 66d 64%
Z Rotation =
0.0 Y Translation =
0 Optical Disc # = Dk 3
Tape Cal # = tape035A.cal0l
MON 2:56 MAY-09-91 ID 10 ROW 20 COL 46 Lmrk 1: 340 DIFF 1: 340 DIFF span 5 rot 237 Vert a of Scans
=
57 X S Y Scale =
1.0 Trig Offset =
31 Filter: Off Points/Scan =
60 SVT
-0.62
.6...................
0.62 36
-0.62 1
- 0.
0 0.62 X Rotation a 60.0 X Translation =
0
+.82v 15Zd Z Rotation =
330.0 Y Translation =
0 Optical Disc # = Dk 3
Tape Cal # = tape035A.cal0l