ML13330B273

1988 Steam Generator Insp Results

ML13330B273
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Site:	San Onofre
Issue date:	03/31/1988
From:	Southern California Edison Co
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1988 STEAM GENERATOR INSPECTION RESULTS SAN ONOFRE UNIT 1 DOCKET NO. 50-206 MARCH 1988 SOUTHERN CALIFORNIA EDISON COMPANY ROSEMEAD, CALIFORNIA 8803:300102 880325 PDR ADOCK 05000206 p

DCD

Table of Contents Section Page I. Introduction 1

II. Technical Specification Inspection 2

A. Introduction 2

B. Steam Generator Leak Test 2

C. Denting Inspection 4

D. AVB Inspection 5

E. General Inspection 5

F. Summary/Conclusion 7

III.

NRC Bulletin 88-02 Inspection 8

A. Introduction 8

B. Results 8

C. Conclusion 8

IV. Roll Transition Zone Cracking Inspection 9

A. Introduction 9

B. Results 9

C. Conclusions 9

V. Cold Leg Top of the Tubesheet Indication Growth 10 Inspection A. Introduction 10 B..

Results 10 C. Conclusion 10 VI. Wrapper Support Bar Inspection 11 A. Introduction 11 B. Results 11 C. Conclusion 11 VII. Steam Generator Inspection Summary and Conclusions 12 A. Summary of Results 12 B. Conclusions 12 VIII. References 13 Appendix A - Tube Gauging Results Appendix B - SG-A, SG-B, and SG-C Inspection Tubesheet Maps

I. INTRODUCTION On February 12, 1988, San Onofre Unit 1 began the second mid-cycle maintenance and surveillance outage of the current fuel cycle. As part of this outage, the steam generator tubing was inspected in accordance with the San Onofre Unit 1 Technical Specification 4.16, Inservice Inspection of Steam Generator Tubing. In addition to the Technical Specification inspection, other areas of the steam generator tubing were inspected to provide supplemental information to assess the condition of the steam generators. The purpose of this report is to provide detailed results of the steam generator inspections performed during the outage to facilitate the NRC review of these results and approval of the corrective action taken at San Onofre Unit 1 as.it relates to the inspection of steam generators.

Consistent with the provisions of Technical Specification 4.16, an inspection was performed addressing requirements for random surveillance of the steam generator tubing and for special surveillances of anti-vibration bars (AVB) wear, progression of denting, and previously detected tube degradation. The last such inspection was conducted in December 1985. In addition, three special inspections were conducted to enhance the assessment of the San Onofre Unit 1 steam generator tubing condition. These three inspections consisted of a roll transition zone cracking inspection, a cold leg top of the tubesheet indication growth inspection and, in accordance with Reference 1, an inspection to address rapidly propagating fatigue cracks in steam generator tubing. Further, in accordance with Reference 2, a secondary side inspection was conducted to visually inspect the intact wrapper support bars. The purpose of this inspection was to verify the bars.had remained intact during operation and did not require removal.

Section II of this report contains the Technical Specification inspection program description, results, corrective actions, and conclusions.Section III of this report contains the description, results, and conclusions of the NRC Bulletin 88-02 inspection.

Section IV contains the description, results, corrective actions, and conclusions of the roll transition zone cracking inspection.Section V contains the cold leg top of the tubesheet indication growth inspection description, results, corrective actions, and conclusions.Section VI contains the wrapper support bar inspection description, results, and conclusions.Section VII summarizes the overall conclusions derived from the inspection program. Finally,Section VIII provides a listing of the references used in this report.

II. TECHNICAL SPECIFICATION INSPECTION A. Introduction The San Onofre Unit 1 Technical Specification steam generator tubing inspection was performed during February 22, 1988, through March 7, 1988. The previous technical specification inspection was performed in December 1985. The December 1985 and earlier inspection results indicated that the pattern of denting in steam generators "A" and "C" is unchanged, and in all other respects the steam generators "A", "B", and "C" (SG-A, SG-B, and SG-C) are behaving in a like manner. Based on Technical Specification 4.16.A.3, which allows the inspection of steam generators on a rotating schedule if they are performing in a like manner, SG-C was selected for the general inspection, special inspection for AVB wear, and special inspection for denting. The inspection plans, results, and conclusions of each of these inspections are discussed below.

B. Steam Generator Leak Test

1. Description San Onofre Unit 1 was experiencing a steam generator primary to secondary leak of approximately 70 gallons per day (gpd) when the plant shut down on February 12, 1988. This leakage had been detected during the current fuel cycle (Cycle 9) and slowly increased to 70 gpd before shutdown. Since this leakage had been detected during the fuel cycle, a leak test had been planned to identify and remove the leaking tubes from service.

2. Results The secondary side of all three steam generators were pressurized to approximately 750 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 twenty leaking tubes were identified in the hot leg channel heads.

The breakdown of leaking tubes per steam generator is shown below:

Tube Number Leak Rate S/G Row Col Tube Identification (Drops/Minute)

A 11 60 Braze Converted Sleeve 0.3 32 69 Mechanical Sleeve

<0.2 32 71 Mechanical Sleeve

<0.2 34 66 Mechanical Sleeve

<0.2 (continued)

Tube Number Leak Rate S/G Row Col Tube Identification (Drops/Minute)

B 11 38 Braze Converted Sleeve 0.5 11 49

-No Braze Converted Sleeve 1.0 17 29 Braze Converted Sleeve

<0.2 26 67 Braze Converted Sleeve 0.3 27 24 Mechanical Sleeve

<0.2 28 30 Mechanical Sleeve

<0.2 36 60 Mechanical Sleeve

>120.0 38 50 Mechanical Sleeve 8.0 38 55 Mechanical Sleeve 7.0 C

1 98 Explosive Plug 3.0 12 27 Braze Converted Sleeve 5.0 12 56 Mechanical Sleeve

<0.2 12 62 Braze Converted Sleeve 0.3 34 52 Mechanical Sleeve 0.4 35 43 Mechanical Sleeve 2.0 37 51 Mechanical Sleeve Steady Stream Each leaking sleeve was inspected with eddy current testing and the results showed that five sleeves (SG-B R11C38, R38C55, R36C60; SG-C R35C43, R34C52) did not have the required upper joint hard roll, and one sleeve (SG-C R37C51) did not have the required lower hard roll.

The leaking sleeve missing the lower hard roll has been removed from service by weld plugging, the leaking explosive plug has been repaired by weld plugging, and the remainder have been removed from service by mechanical plugging.

3. Conclusion Although six of the leaking sleeves were missing hard rolls, all but two (SG-B R36C60 and SG-C R37C51) of the 19 leaking sleeved tubes can be definitively quantified to be within the leakage permitted for each sleeve joint as discussed in Reference 3 (210 drops per minute). Therefore, on the basis that: (1) all leaking sleeves have been removed from service, (2) the leakage experienced from these sleeves increased slowly over the fuel cycle without exceeding technical specification limits, and (3) the progression of IGA in the sleeved tubes is projected to be less than 1% per month based on Reference 4; it has been concluded that operation can continue through the end of the current cycle (approximately 3 effective full power months). Prior to restart from the refueling outage at the end of this cycle all of the baseline and subsequent eddy current data for the San Onofre Unit 1 sleeves will be reevaluated to

determine the presence of sleeve expansions and hard rolls.

Based on this evaluation, and a more detailed evaluation of the effects of operating with sleeves without expansions/hard rolls, required corrective actions will be taken during the next refueling outage.

C. Denting Inspection

1. Description As a result of previous steam generator inspections, 190 tubes in SG-C hot and cold legs were identified as having restrictions due to steam generator denting. These tubes were gauged through their respective restricted support plate. Each tube was gauged using the probe size which was previously recorded as having passed through the support. In the event of an increased restriction, the tube was gauged with successively smaller probes until a probe passed the restriction.

Restriction sizes observed this outage were compared to previous inspection results to assess the progression of steam generator tube denting.

2. Results The results of the tube gauging were compared with the corresponding data from previous inspections for the 190 tubes. The comparison identified one tube which restricted a probe size that previously passed and six tubes which passed a probe size that was previously restricted. A table showing the results of these seven tubes is provided in Appendix A. The remaining 183 tubes gauged remained unchanged, passing the probe size which was previously passed.

3. Conclusion Out of 190 tubes gauged, only one tube was more restricted than in previous inspections. This result is within the expected variability of the gauging process and is not attributable to a significant progression of the denting process. In addition, Reference 5 evaluated the denting phenomena at San Onofre Unit 1 and concluded that denting has not progressed in the Unit 1 steam generators since 1972 and the required special dent gauging inspections could be eliminated. Therefore, based on the inspection results and the evaluation in Reference 5, steam generator tube denting is not an active phenomena occurring at San Onofre Unit 1.

D. AVB Inspection

1. Description All tubes in SG-C with greater than 20% AVB indications were inspected. This inspection fully incorporated those tubes which were required to be tested by Technical Specification 4.16.C.l. A total of 240 tubes with greater than 20% AVB indications were inspected.

2. Results The results of the SG-C AVB inspection were compared to previous inspection results and the AVB indications exhibited apparent growth. Due to this apparent growth, approximately 98% of the tubes previously tested in 1985 had their 1988 DDA-4 outputs compared to the 1985 DDA-4 output. This comparison was performed to determine whether they represented actual changes in AVB related degradation or were simply artifacts of the examination and analysis techniques utilized. Based upon this comparison, it was determined that the results were artifacts of the examination process, and, therefore tube degradation at the AVBs has not progressed.

3. Conclusion Based on the comparison performed, it was concluded that tube degradation at the AVBs has not progressed and the discrepancies in measured values are attributable to variations in measurement techniques from the 1985 examination to the 1988 examination.

E.

al In

1. Description The general inspection program consisted of inspecting the non-sleeved length of 317 steam generator tubes in SG-C (at least 3% of the total number of tubes in service). In addition, the sleeved portion of 277 tubes (approximately 4% of the total sleeved tubes) was inspected. Two pitches around the sleeving boundary and a random pattern in the remaining peripheral tubes (452 tubes) were inspected to identify any new indications of intergranular attack (IGA).

Further, approximately 5,949 additional tubes were inspected in all three steam generators to address three specific concerns.

These concerns were rapidly propagating fatigue cracking (NRC Bulletin 88-02), roll transition zone cracking, and growth at the top of the cold leg tubesheet. These specific programs are discussed in Sections III, IV, and V, respectively. 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 8x1 and the motorized rotating pancake coil (MRPC) probe were utilized to supplement the bobbin probe when necessary. The magnetically biased bobbin probe and the crosswound probe were employed to assess the condition of the sleeved tubes inspected.

2. Results As a result of the general technical specification and additional eddy current testing program, a total of 148 tubes were required to be removed from service. The breakdown of these pluggable tubes is 7, 59, and 82 tubes in SG-A, SG-B, and SG-C respectively. A total of 147 tubes were plugged due to hot leg tube roll transition zone cracking and 1 was plugged in SG-C due to a 62% indication just above the hot leg tubesheet.

In addition to the 148 that required plugging, during this outage three tubes were preventively mechanically plugged due to degradation quantified to be just under the plugging limit.

The requirement for eddy current testing program expansion in accordance with Technical Specification 4.16.B.2 was satisfied by inspecting 100% of the non-sleeved portion of the hot leg in all three steam generators to address the roll transition zone cracking which is further discussed in Section IV of this report.

A limited amount (less than 10% of the tubes inspected) of new greater than 20% indications and more than 10% growth was indicated at either the top of the cold leg tubesheet, the top of the hot leg tubesheet, and the first and second cold leg tube supports. In accordance with Technical Specification 4.16.8.1, the above conditions required a 3% expansion into one of the uninspected steam generators. However, due to the eddy current inspections planned to address the three previously mentioned problems, more than 3% of the tubes in the other two steam generators were inspected. In no case did the results of these inspections meet the growth expansion criteria in Technical Specification 4.16.B.2. Therefore, no further expansions were required.

3. Conclusions The tubes selected for this inspection included random tubes and tubes in critical areas identified by Unit 1 and other similar plant experience. All tubes required to be removed from service, based on the eddy current testing results, were mechanically plugged. Current Technical Specification requirements to inspect previously identified problem regions during future inspections will ensure corrective actions are performed as necessary to prevent potential tube failures.

The two tubes with IGA indications (SG-B R26C21 and SG-C R33C76) were evaluated and compared to previous results to assess if IGA was progressing. Tube R33C76 had not be tested since 1980 and the IGA indication was identified this inspection utilizing improved examination and analysis techniques. Tube R26C21 was evaluated, during the 1985 inspection, as having a distorted signal at the top of the hot leg tubesheet and additional inspection this outage with the MRPC, confirmed that it was an IGA indication. Neither of these IGA indications were above the 50% plugging level by evaluation of the bobbin coil 100 KHz absolute data. Both tubes were preventively plugged. Based on this evaluation it is concluded that IGA is not progressing at San Onofure Unit 1.

F. Summary/Conclusion A total of 7,425 tubes (72 percent of the tubes in service) were inspected and 169 tubes were removed from service by mechanical plugging. In addition, I sleeved tube was removed from service by weld-plugging and 1 previously plugged tube was repaired by weld plugging. The tubesheet maps showing the tubes inspected in SG-A, SG-B, and SG-C are shown in Appendix B. These maps show all tubes inspected during this outage. The 169 tubes which required plugging included 147 defects in roll transition zone on the hot leg side, 3 defects at or above the top of the hot leg tubesheet, and 19 leaking sleeves.

The 147 defects were identified and plugged as a result of improved inspection and analysis techniques. This inspection has demonstrated that there has been no detectable progression of IGA, denting, AVB wear, or sleeve degradation. Accordingly, it is concluded that the remedial action taken (plugging) is appropriate to resolve steam generator tube degradation identified during this inspection and no further action is required.

III.

NRC BULLETIN 88-02 INSPECTION A. Introduction As the result of a tube rupture at the North Anna nuclear power plant, Reference 1 was sent to the owners of all Westinghouse nuclear power plants requesting information regarding the condition of the tube/support plate intersection at the fourth support plate and the exact location of the AVBs. This information was to be analyzed to determine the susceptibility of the San Onofre Unit 1 steam generator tubing to rapidly propagating fatigue cracking.

B. Results A total of 1,796 tubes were inspected in all three steam generators to assess the condition of the susceptible tubes. This inspection.

consisted of inspecting 100% of the in service tubes in Rows 11 through 17 in all three steam generators. The data for each tube is being analyzed for the presence of denting (magnetite) at the hot and cold leg fourth support plate and for the location of its intersection with the associated AVBs.

C. Conclusion The analysis of the data obtained is in progress. The results of this evaluation will be submitted separately as required in Reference 1.

IV. ROLL TRANSITION ZONE CRACKING INSPECTION A. Introduction During preparation for the San Onofre Unit 1 steam generator eddy current testing program, the eddy current testing contractor (Conam) was reviewing previous eddy current data. During this review, it was discovered in late January 1988 that distorted indications in the hot leg roll transition zone, similar to those recently found at the Connecticut Yankee nuclear power plant, existed at San Onofre Unit 1. Specifically, data of approximately 340 tubes from the past inspection was reviewed and indications were observed in the tube roll transition zone, but no indications were observed in the tube roll expansion. Therefore, to address the possibility of roll transition zone cracking at San Onofre Unit 1, an examination of 100% of the hot leg non-sleeved tubes in each steam generator was conducted. Additionally, included in the cold leg inspection described in Section V of this report and as noted in References 6 and 9, a sample of cold leg tubes was inspected in the tube roll expansion region.

B. Results A total of 3,912 hot leg roll transitions were examined and evaluated. Each transition that had any anomalous indication was examined with a MRPC probe to determine if the tube should be removed from service. A total of 147 tubes were removed from service by mechanical plugging due to roll transition zone cracking. However, as part of these detailed inspections, using probes and techniques specifically designed to inspect the roll transition zone, it was discovered that the use of the F* criterion would benefit San Onofre Unit 1. Accordingly, the F* criterion (Reference 6) was utilized to leave 44 tubes in service which otherwise would have been plugged. The results for the inspection of the cold leg tubes indicated that this problem is not present in the cold leg side.

C. Conclusions A comparison of the 1985 and 1988 eddy current data was conducted to determine if the roll transition zone cracking was active. This comparison showed a slight change in the vertical distortion in only 3 out of the 13 tubes compared. Further, none of the approximately 340 tubes which had previously been inspected with the MIZ-18 showed any change in the eddy current signal. In future inspections, the requirement established in Reference 9 for continued monitoring of degraded tubes will ensure that roll transition cracking will be monitored for growth.

V. COLD LEG TOP OF THE TUBESHEET INDICATION GROWTH INSPECTION A. Introduction As a result of the discovery of a limited amount of indication growth at the top of the cold leg tubesheet during the 1985 inspection, an inspection of previously degraded and random cold leg tubes in all three steam generators was conducted to monitor this growth.

B. Results A total of 868 tubes were inspected to the first cold leg support to monitor top of the tubesheet growth in all three steam generators. The results of this inspection showed that there was limited indication growth in this region (less than 10% of the tubes inspected were affected) and required expansion into another uninspected steam generator as per Technical Inspection 4.16.8.1.

This expansion was incorporated into the planned inspection in all three steam generators. At no time did the results of this inspection meet the expansion requirements of Technical Specification 4.16.B.2.

C. Conclusion The inspection of these cold leg tubes was to monitor the growth of indications found in the 1985 inspection. No tubes inspected required plugging. However, due to the limited growth in this region, existing requirements to inspect previously identified problem regions during future inspections will ensure that this area continues to be inspected and corrective actions are performed as necessary to prevent potential failures.

VI. WRAPPER SUPPORT BAR INSPECTION A. Introduction The original wrapper support design for the Westinghouse Series 27 steam generator, including San Onofre Unit 1, included six symmetrically located and vertically positioned bars welded to the base of the wrapper on the ID and threaded into the tubesheet. The wrapper rested on these bars and the bars were intended to accept the vertical wrapper loads specified in the steam generator equipment specification.

Subsequent modifications to Series 27 steam generators involved installing two brackets (Type I) in each steam generator, one end of the bracket welded to the transition section of the upper portion of the wrapper assembly with the other end attached to the feedwater ring bracket close to the steam generator shell.

These brackets were designed to prevent vertical displacement of the wrapper assembly even if all of the existing wrapper support bars were not in place. In order to provide further support to the wrapper, these two support brackets were supplemented by a third bracket (Type II) welded to the wrapper and attached to the feedwater ring nozzle support.

During the secondary side visual inspections conducted in 1982, all but three of the wrapper support bars were found to be either broken or missing. The subsequent investigation required the loose support bars to be removed but allowed the three intact support bars to remain.

In Reference 2, a commitment was made to visually inspect the intact wrapper support bars in SG-A and SG-B during the next refueling outage.

B. Results A visual inspection of the intact wrapper support bars was conducted. The results of the inspection showed that the support bars are still intact and have not moved.

C. Conclusion Based on the results of the wrapper support bar investigation documented in Reference 7 and the fact that the wrapper support bars in SG-A and SG-B remain intact, the wrapper support bars can be left in place without affecting tube integrity.

VII.

STEAM GENERATOR INSPECTION

SUMMARY

AND CONCLUSIONS A. Summary of Results As part of the inservice inspection of steam generator tubes, a total of 7,425 tubes were inspected. The inspection results indicate no detectable progression of IGA, denting or AVB wear.

Further, there was no detectable sleeve degradation. Although 170 tubes were plugged, 147 of these were due to roll transition zone cracking and required plugging due to improved inspection and analysis techniques, 19 were due to primary to secondary leakage and 3 were preventively plugged. In conclusion, only one tube required plugging due to indication growth beyond the plugging limit. The tubes plugged this outage increases the total equivalent plugging level to 13.8%. This plugging level is consistent with the current analyzed plugging level of 20%

addressed in Reference 7.

The wrapper support bar visual inspection results demonstrate that the three remaining wrapper support bars in SG-A and SG-B are intact.

B. Conclusions The information provided in Sections II, III, IV, and V of this report establishes the basis for concluding the remedial action taken to resolve the steam generator tube degradation identified during this inspection is appropriate. Accordingly, no further action is required and power operation can be safely resumed.

In regards to the wrapper support bars, the information provided in Section VI of this report and Reference 7 provides adequate basis for leaving the intact bars in SG-A and SG-B. To ensure these support bars remain intact, an inspection will be conducted during the next refueling outage.

In summary, the information presented in this report provides adequate basis for the approval of the corrective action taken at San Onofre Unit I as it relates to Technical Specification inspection of steam generator tubing.

VIII. REFERENCES

1. NRC Bulletin No. 88-02:

"Rapidly Propagating Fatigue Cracks in Steam Generator Tubes," dated February 5, 1988

2. Letter, M. 0. Medford (SCE) to G. E. Lear (NRC), "Steam Generator Inspection Report," dated April 14, 1986

3. "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

4. "1985 Re-Evaluation of Steam Generator Inspection Interval, San Onoffre Nuclear Generating Station Unit 1" dated March 1985, submitted by letter, M. 0. Medford (SCE) to 3. A. Zwolinski (NRC),

March 19, 1985

5. Letter, Kenneth P. Baskin (SCE) to USNRC, "Amendment Application No. 144," dated October 30, 1987

6. Letter, Kenneth P. Baskin (SCE) to USNRC (NRC), "Amendment Application No. 149," dated March 10, 1988

7. Letter, K. P. Baskin (SCE) to D. M. Crutchfield (NRC), "Steam Generator Inspection Report," dated September 12, 1982

8. Letter, D. M. Crutchfield (NRC) to R. Dietch (SCE), "Steam Generator Repair Program and Plant Restart," dated June 8, 1981

9. Letter, M. 0. Medford (SCE) to USNRC (NRC), "Revision to Proposed Change No. 182", dated March 22, 1988.

9458F

• e4.

APPENDIX A TUBE GAUGING RESULTS RESTRICTED TUBE DATA SAN ONOFRE UNIT 1 STEAM GENERATOR 'C' INLET Tube Number Probe Size Probe Size (Inches)

(Inches)

Passed Passed Row Column 1985 1988 1

19

.500

.560 1

36

.500

.560 1

49

.500

.560 1

62

.560

.500 3

50

.460

.500 8

2

.500

.560 9

2

.500

.560 APPENDIX B SG-A, SG-B, AND SG-C INSPECTION TUBESHEET MAPS 02/8B, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT 1 STEAM GENERATOR: A DATE: 03/06/88 PROC: SO1-XXVII-3.1 TIME:

14:05:38 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM HOT LEG

.A so

3.

5 20

-M-M

-REVAL.HOT 1206 112 2

~

"PLUGGED 381 100 95 90 85 80 75 70 65 60 55 50 45 40 35 30 25 20 15 10 5

CONAM

02/BB, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: A DATE: 03/06/88 PROC: SOI-XXVII-3.1 TIME:

14: 31: 44 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM COLD LEG 40 35

-30 25 10 EVAL.

COLD 965 1LJR "II M" ""

PLUGGED 381 100 95 90 85 60 75 70 65 60 55 50 45 40 35 30 25 20 15 10 5

CONPAM

02/88, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: B DATE: 03/07/88 PROC: SOI-XXVII-3.1 TIME:

I1: 47:35 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM HOT LEG 35 HOT 1358 PLGGD00 jil!,

ii iifIIIi il Iil 1

u fI iihj 111111

.1.

IlfI l iijl i il t I

-144.5 EVAL HO 1358.J 10095 90 85 80 75 70 65 60 55 50 45 40 35 30 25 20 15 10 5

00NAM

02BSOUTHERN.CALIFORNIA EDISON,, SAN ON0FRE, UNIT I STEAM GENERATOR: B DATE: 03/07/88 PROC: SOI-XXVII-3.1 TIME: 11:17:48 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM COLD LEG

.!IJ-I I

I

.45 11

[

4 0 I

I II I35 I~~~~VL COL 011 1111 I l l i l l l l l l i 1 1 1

1 1 1 1 1 1 I I I l I l l 11 1 1 i l I I I 1 1 1 1 1 1 1 I I I I I 1 1 1 1 1

I I

10 95 90 85 80 75 701 65 60 55 50 45 10 35 3

25 0

15 105 T

I II I I I ICONAM

02/88, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: C DATE: 03/06/88 PROC: SOI-XXVII-3. 1 TIME:

15: 56: 12 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM HOT LEG 45 30 EVAL.

HOT 1348

-- m

-0 PLUGGED 364 100 95 90 85 80 75 70 65 60 55 50 45 40 35 30 25 20 15 10 5

CONAM

02/BB, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT I STEAM GENERATOR: C DATE: 03/06/88 PROC: SOI-XXVII-3.1 TIME:

16:19:49 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM COLD LEG 45 40 ZZ 30 4

25 20

-EVAL.

COLD 1260 PLUGGED 364 1111 il li 111 III I if 11 1 11 1111 1111 I IIIIf I I I l i 1111 i i fII111 1111!III IlliIll Ill ll I I 1111 100 95 90 85 80 75 70 65 60 55

.50 45 40 35 30 25 20 15 10 5

CONAM

02/88, SOUTHERN CALIFORNIA EDISON, SAN ONOFRE, UNIT STEAM GENERATOR: C DATE: 03/06/88 PROC: SOi-XXVII-3.1 TIME:

18: 11: 19 LOCATION: ALL CRITERIA: TUBES EVALUATED FROM HOT LEG 45 40 PLGGE 35 4-~~--

25

-LT 11~~

10 14 100 95 90 85 80 75 70 65 60 55 50 45 40 35 30 25 20 15 1.0 5

CONAM