IR 05000498/1993028

Insp Repts 50-498/93-28 & 50-499/93-28 on 930810-12,23-27 & 0929-1005.No Violations Noted.Major Areas Inspected:History & Matl Condition of Units 1 & 2 Steam Generator Tubing & to Assess Effectiveness in Detecting & Correction of Tubing

ML20059J909
Person / Time
Site:	South Texas
Issue date:	10/20/1993
From:	Thomas C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20059J893	List: IR 05000498/1993028 ML20059J891
References
50-498-93-28, 50-499-93-28, NUDOCS 9311150100
Download: ML20059J909 (22)

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APPENDIX U.S. HUCLEAR REGULATORY COMMISSION

REGION IV

Inspection Report: 50-498/93-28 50-499/93-28 Licenses: HPF-76 NPF-80 Licensee: Houston Lighting & Power Company P.O. Box 1700 Houston, Texas Facility Name: South Texas Project Electric Generating Station, Units I and 2 Inspection At: Wadsworth, Texas Inspection Conducted: August 10-12 and 23-27, and September 29 through October 5, 1993 Inspectors: 1. Barnes, Technical Assistant, Division of Reactor Safety L. Gilbert, Reactor Inspector, Maintenance Section, Division of Reactor Safety Accompanying personnel: W. Sifre, Reactor Engineer, Division of Reactor Projects Dr. C. Dodd, Consultant, Oak Ridge National Laboratory

Approved: [ Seh6f O/kmw&v fo/pp' /LG Cecil 0. Thomas,~ Acting Deputy Director Date Division of Reactor Safety Insnection Summary Areas Insnected (Units 1 and 2)t Regional initiative, announced inspection to review the history and material condition of Units 1 and 2 steam generator tubing, and to assess the effectiveness of licensee programs in detection and analysis of degraded tubing, repair of defects, and correction of conditions contributing to tube degradatio PDR ADOCK 0500

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-2-i Results (Units 1 and 21:

• Operational experience is limited since South Texas Project, Units 1-and 2, are the only U.S. pressurized water reactors which utilize Westinghouse Model E steam generators in the plant' design (Section 2.5).

• These units have been operated with a hot leg temperature of 626*F, which appeared from available information to be the highest temperature used by any domestic pressurized water reactor. It was noted by the inspectors that reduction of hot leg temperature.is being pursued by other individual licensees, including South _ Texas Project, as an approach to limit initiation and propagation of stress corrosion cracking (Section 2.1).

• Actions were taken by the licensee to minimize tubing wear in the preheater section of the steam generators by expanding the tubes at two baffle plate locations; and actions were taken to improve resistance to stress corrosion cracking by peening of tube expansion transition areas and heat treatment of low radius U-bends (Section 2.6).

• Belgian operating expo ience data provided by the licensee indicated that significant stre a corrosion cracking damage had occurred in their Model E steam generators since commercial operation began in 1985 (Section 2.5).

• The 1993 eddy current examination results for South Texas Project, Units 1 and 2, indicated that limited tube degradation had occurred in Unit 1. Similar damage indicatione were not identified in Unit 2 tubing. Tube pull satples will A hjected to laboratory examination '

to verify whether tube degradation uc occurred and the nature, as applicable, of the damage mechanisms ($ection 2.4). '

The licensee adopted a comprehensive eddy current examination strategy for the current steam generator examinations. With one exception, prior inservice examinations were performed using caly the bobbin method and a sample size at or near the minimum required by the Technical Specifications (Section 4.1).

• The current eddy current examination program requirements were found to be good, with the primary area of improvement being the adoption of'

formalized training and testing of data analysts (Section 4.2). :

• The 1993 eddy current data were observed to exhibit low noise, with the performance of the contractor analysts being found to be satisfactory ,

for the tube data sample that was reviewed (Section 4.3).

• Visual examination of Unit 2 steam generators appeared to have been well performed for the documented inspection scope. Procedural guidance !

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-3-lacked specificity, however, on inspection scope expectations (Section 3.1). t

• Since commercial operation of STP, Units _1 and 2,'the secondary water i chemistry program for both units had continually been upgraded to !

incorporate industry guidelines as they were made available (Section 5.1). ,

• The licensee has maintained excellent control of the secondary water i chemistry, with only _two significant out-of-specification chemistry conditions noted since plant startup. These conditions both involved out-of-specification sodium concentrations that occurred in Unit I during 1990 and again in 1993. In each case, the out-of-specification condition was promptly-identified and corrected (Section 5.5). .

Summary of Inspection Findinos:

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• No inspection findings were opened or close Attachment:

• Attachment - Persons Contacted and Exit Meetirig j i

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-4-DETAILS 1 STEAM GENERATOR TUBE INTEGRITY REVIEW (73755, 79501, 79502)

The objectives of this inspection were: (a) to ascertain the history and-material condition of the Units 1 and 2 steam generator tubing; and (b) to assess the effectiveness of licensee programs in detection and analysis of degraded tubing, repair of defects, and correction of conditions contributing to tube degradatio STEAM GENERATOR MATERIALS AND TUBE DEGRADATION HISTORY

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2.1 Steam Generator Description

South Texas Project Electric Generation Station (STP), Units 1 and 2, are Westinghouse-designed 1250 megawatt electric pressurized water reactors, which commenced commercial operation in March 1988 (Unit 1) and June 1989 (Unit 2).

The STP unit design utilized four Westinghouse Model E recirculating steam generators. This model of steam generator contained a preheater section and utilized Inconel 500 (ASME Material Specification SB-163) U-tubes with a nominal diameter and wall thickness, respectively, of 0.75 inches and 0.043 inches. The 7 umber of tubes in a steam generator differed slightly between units (i.e., 4,864 in Unit I and 4,851 in Unit 2). Secondary side

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tube support structures consisted of 10 drilled tube support plates on the ,

inlet side,13 drilled tube support and baffle plates on the outlet side, and anti-vibration bars which-intersected tubes in up to 4 locations in the U-bend area. Unit I tube support and baffle plates were fabricated from carbon steel, whereas the Unit 2 plates were fabricated from Type 405 ferritic stainless steel. With the exception of the 1.12-inch thick top support plate in Unit 2 steam generators, the tube support and baffle plate thickness used was 0.75 inches. The inspectors considered that the difference in tube support materials would make the Unit I steam generators more susceptible than those in Unit 2 to long-term tube denting and stress corrosion damage due to the presence of corrosion products and magnetite formation in the carbon stee The inspectors ascertained that a steam generator primary side inlet temperature of 626'F was used in unit operation. The inspectors noted that based on available industry information, this placed the STP design at the top of domestic pressurized water reactors with respect to highest hot leg temperature. It was noted by the inspectors that reduction of hot leg temperature is being pursued by other individual licensees as an approach to limit initiation and propagation of stress corrosion crackin ;

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2.2 Tubino Material The inspectors reviewed the data packages for the steam generatore. installed ,

in STP, Units 1 and 2. The data packages included the ASME ManuScturers Data Reports, steam generator major parts list, and certified materiai ' et reports

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-5-for each steam generator. The ASME Manufacturers Data Reports listed the applicable ASME Code asSection III,1974 Edition and Code Case 1484, with the material for the steam generator tubing specified as ASME Materials Specification SB-163. The data packages additionally referenced Westinghouse Specification 2656A84, " Material-Nickel-Chromium-Iron Tubing (High Yield Strength per Code Case 1484)," as being applicable to the steam generator tubing material. The inspectors requested that this document be obtained for review. The document was subsequently furnished by Westinghouse and was orally indicated to be considered proprietary, although not marked as suc The steam generator major parts lists identified the tubing material used in each steam generator by test numbers. The documentation identified by the test numbers included the certified material test report data for the chemical

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analysis, tensile and hardness testing, flare test, ultrasonic examination, eddy current testing, and hydrostatic testing for each heat of material used to manufacture the steam generator tubes. A random sample of five heat numbers from each steam generator was selected for review of the test dat ,

The heat numbers reviewed are listed belo * Unit 1, Steam Generator A-Test No. T06004, Heats 7342, 7398, 7507, 7508, and 7551

• Unit 1, Steam Generator B-Test No. 105633, Heats 8440, 8442, 8443, 8444, and 8464

• Unit 1, Staam Generator C-Test No. T06013, Heats 8389, 8392, 8395, 8407, ,

and 8410

• Unit 1, Steam Generator D-Test No. T06154, Heats 8714, 8718, 8740, 8755, and 8775

• Unit 2, Steam Generator A-Test No. T08353, Heats 0005, 0013, 0026, 0065, and 0076

lnit 2, Steam Generator B-Test No. T07798, Heats 8901, 8912, 8915, 8916, -

and 8925 *

Unit 2, Steam Generator C-Test No. T06221, Heats 8523, 8529, 8534, 8535, and 8538

Unit 2, Steam Generator D-Test No. T07799, Heats 9099, 9103, 9104, 9108, and 9109 The chemical analysis and mechanical test results were found by the inspectors to be in conformance with the requirements of ASME Material Specification SB-163, Westinghouse Specification' 2656A84, and ASME Code Case 1484. The inspectors noted that the material specifications required the tubing to be supplied in the annealed condition, but did not specify the annealing temperature to be used for the ASME SB-163 tubes. The inspectors were unable to verify the actual temperatures since the certified material l

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-6-test reports furnished by the tubing manufacturer did not indicate the actual-annealing temperature used. The material test reports documented that the tubing had been hydrostatically tested at 3106 psig, minimum. No anomalies or excessive range of reported mechanical properties were identified during review of the tubing certified material test report .3 Tube-To-Tube Sheet Ex3ansion The inspectors were informed by licensee personnel that tubes were expanded after insertion into the tube sheet (i.e., the forging used to support the U-tubes) drilled holes by mechanical rolling for Unit I and hydraulic expansion for Unit 2. The inspectors requested that the applicable tube-to-tube sheet expansion procedures ~be obtained for review. These documents were furnished by Westinghouse and were also orally indicated to be considered proprietary, although not marked as such. The inspectors noted from review of '

the two documents (i.e., Process Specification 81007 JA, " Full Depth Rolling of Steam Generator Tubes," effective December 8, 1977, and Process Specification 81013 RM, " Hydraulic Tube Expansion," effective July 24,1981)

that both processes provided for full-depth expansion of the tubes in the tube sheet. The inspector noted that the inspection verification requirements were more detailed for the hydraulic expansion proces .4 Steam Generator Tube Deoradation History 2. Unit 1 Prior to operational service, Unit I steam generators contained a total of 15 plugged tubes (i.e., Steam Generator A - 3, Steam Generator B - 4, Steam Generator C - 3, and Steam Generator D - 5). Eddy current examinations were performed on a sample of steam generator tubes during Refueling Outages IRE 01 (Fall 1989) and IRE 03 (Spring 1991), with no repairable indications foun These refueling outages occurred, respectively, after 0.82 and 1.55 effective full-power years (EFPYs) of operatio As of this inspection, Unit I had accrued 2.88 EFPYs of operation. The status of the Unit 1 examinations was obtained during the final onsite portion of the_-

inspection on September 29-30, 1993. At that time, bobbin coil examination *

had identified a repairable indication in one tube in Steam Generator C (i.e.,

wear had exceeded 40 percent throughwall at an anti-vibration bar area).

Motorized rotating pancake coil (MRPC) examinations of a random sample of 21 percent of the tubes (see Section 4 for additional details) in each steam generator had identified the presence of multiple axial indications in one ,

tube in both Steam Generators A and B. These indications were reported after initial rtview to potentially be inside diameter initiated cracks which were located, respectively, just below the tube sheet secondary face in Steam Generator A and at approximately 1 inch below the tube sheet secondary face in Steam Generator B, In addition, MRPC examinations had identified three potential single axial cracks and one potential circumferential crack (i.e.,

Steam Generator B - one single axial indication, Steam Generator C - two single axial indications, Steam Generator D - one circumferential indication).

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-7-The circumferential indication was located in the tube expansion transition area at the outside diameter of the tube. One of the Steam Generator C single axial tube indications was located at the.first tube support plate on the hot leg side at the outside diameter of the tube. The remaining two axial indications were located at the inside diameter of the tube '

As a result of the MRPC detected indications, the licensee expanded the scope of the MRPC examinations to 100 percent of the tubes at the top of the tube sheet. The augmented examinations had been completed for Steam Generator D by September 30, 1993, with an additional two multiple axial indications-and one circumferential indication being found. The location of these indications was inside diameter for the axial indications and outside diameter for the -

circumferential indication. The licensee noted that the bobbin coil examinations had not detected either of the circumferential indications' or one of the multiple axial indications that was detected during the 100 percent MRPC cxamination of Steam Generator D at the top of the tube' sheet. The inspectors were informed on October 8, 1993, by telephone that the completed MRPC examinations had identified an additional six outside diameter circumferential indications at the top of the tube sheet in Steam Generator Licensee personnel also indicated that tube pulls would be made to allow laboratory examination of the defect indications and, thus, provide *

confirmation of operative degradation mechanism .4.2 Unit 2 Prior to operational service, Unit 2 steam generators contained a total of 66 plugged tubes (i.e., Steam Generator A - 16, Steam Generator B - 13, Steam Generator C - 22, and Steam Generator D - 15). Eddy current examinations were performed on a sample of steam generator tubes during Refueling Outages 2RE01 (Fall 1990) and 2RE02 (Fall 1991), with no repairable indications found. One tube was preventively plugged during Refueling Outage 2RE02 as a result of the identification of 38 percent throughwall wear at a contact area with an anti-vibration bar. These refueling outages occurred, respectively, after.0.84 and 1.55 EFPYs of operation. As of this inspection, Unit 2 had accrued 2.64 EFPYs of operation. Comprehensive eddy current examinations (see Section 4 for_ ;

additional details) were in progress on all four Unit 2-steam generators during the August 23-27, 1993, portion of the inspectio At the end of this portion of the inspection, analysis was approximately 70 percent and 80 percent complete, respectively, for planned bobbin coil and MRPC examinations. The preliminary results did not indicate that any significant amount of degradation had occurred in the steam generator tubin Only one repairable indication had been identified as of August 27, 199 This indication was found in Steam Generator A and was determined to be anti- '

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vibration bar wear. Examination by MRPC identified an axial indication in a Row 2 low radius U-bend (i.e., Tube 2-49) at a location where preservice examination had previously identified a manufacturing buff mark. In order to more definitively determine whether the axial indication represented a crack, ;

the licensee procured a special MRPC probe to further examine the indicatio This probe used three coils, one normal pancake coil, one oriented to detect '

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-8-axial cracks, and one oriented to detect circumferential cracks. The inspectors were informed on August 31, 1993, by telephone that the results from this examination did not indicate that a crack was present. The inspectors were additionally informed by telephone on September 7,1993, that the completed analyses had not identified any additional repairable indications. Two other tubes were preventively plugged'as a result of the identification of anti-vibration bar wear in excess of 30 percen .5 Model E Steam Generator Operational Experience STP, Units 1 and 2, are the only domestic Westinghouse pressurized water reactors which utilize Model E steam generators, thus limiting the available -

Model E operational experience information. The inspectors requested licensee staff to furnish any readily available information they may have in regard to two Belgian pressurized water reactors which utilized Model E steam generators (i.e., Doel 4 and Tihange 3). Belgian information provided by the licensee with respect to Doel 4 indicated that the unit commenced commercial operation in 1985, and that it had experienced stress corrosion cracking (SCC) at different tubing locations during its operational history. Specifically, primary water SCC was first identified at tube sheet locations in 1986, cellular outside diameter SCC at tube support plate locations in 1990, circumferential roll transition area SCC in 1991, and outside diameter axial SCC in free span areas in 1992. Information provided with respect to Tihange 3 indicated that it also commenced commercial operation in 1985, with primary water SCC at tube sheet locations being first identified in 198 Circumferential SCC (inside diameter and outside diameter) was identified in roll transitions in 1991, as was cellular outside diameter SCC at tube support plate location .6 Licensee Actions Taken to increase Tubina SCC Resistance and Minimize Wear The inspectors were informed by licensee personnel that resistance to primary water SCC was increased in steam generator tube expansion transition areas by use of peening techniques to induce surface compressive stresse Specifically, roto peening was performed in these areas in Unit 1 prior to -

unit operation. Shot peening was utilized for Unit 2, with the tube expansion transition areas being peened on the hot leg side during the first refueling outage and on the cold leg side during the second refueling outage. The licensee had acquired Belgian technical literature which indicated that the use of these peening techniques should significantly reduce the incidence of-primary water SCC at these location The inspectors additionally ascertained that heat treatment of the low radius Rows 1 and 2 U-bends had been performed at the STP-site by Westinghouse to increase the resistance of the bend region to SCC. This activity was accomplished during the first outage for Unit I and prior to operation for Unit 2. Review of Westinghouse conformance with the technical requirements of its heat treatment procedure was not performed during the inspectio .

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-9-Modifications were also made by the licensee in the preheater section of the steam generators prior to Units 1 and 2 operation, for the purpose of minimizing tubing wear. The modifications-consisted of expansion of 160 tubes at the "B" and "D" baffle plate location .7 Conclusions

• STP, Units 1 and 2, are the only U.S. pressurized water reactors which utilize Westinghouse Model E steam generators in the plant desig * These units have been operated with a hot leg temperature of 626'F, ,

which appeared from available information to be the highest temperature used by any domestic pressurized water reactor. It was noted by th inspectors that reduction of hot leg temperature is being pursued by individual licensees, including South Texas Project, as an approach to limit initiation and propagation of SC * The most significant differences noted between the Units 1 and 2 steam generators were the Unit I use of carbon steel tube support plates and mechanical rolling of tubes into the tube sheet, versus the Unit 2 use of Type 405 ferritic stainless steel tube support plates and hydraulic expansion of tubes into the tube sheet. The inspectors considered that these differences would make the Unit 2 steam generators-less susceptible than those in Unit I to long-term denting and SCC damag * Belgian operating experience data provided by the licensee indicated that significant SCC damage had occurred in their Model E steam generators since commercial operation began in 198 * The 1993 eddy current results for STP, Units 1 and 2, indicated that limited tube degradation had occurred in Unit 1. Similar damage indications were not identified in Unit 2 tubin * Tube pull samples will be subjected to laboratory examination to verify-whether tube degradation has occurred and the nature, as applicable, of the damage mechanism * Actions were taken by the licensee to minimize tubing wear in the preheater section of the steam generators by expanding the tubes at two baffle plate locations; and actions were taken to improve resistance to '

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stress corrosion cracking by peening of tube expansion transition areas-and heat treatment of low radius U-bend >

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l-10-I 3 VISUAL EXAMINATION OF THE SECONDARY SIDE OF THE UNIT 2 STEAM GENERATORS 3.1 Review of Procram Recuirements and Inspection Data The inspectors reviewed Procedure OPEP07-SG-0002, " Steam Generator Internal Inspection," Revision 0, and the results of secondary side visual inspections performed by B&W Nuclear Technologies (BWNT) during March and April 1993, of the four Unit 2 steam generators. The inspectors noted that the inspection attributes developed by the licensee included observed damage (i.e., erosion, q cracking, denting, pitting), sludge bridging between tubes, tube hole crevice ,

blockage, and the presence of foreign objects. The inspections performed by !

BWNT included for.each steam generator: a general area inspection of the j steam drum, as well as a section of the lower set of anti-vibration bars in the U-bend region; baffle and tube support plates in the preheater region; and J pre- and post-sludge lancing inspections to observe tube sheet sludge condition The steam drum inspections did not identify any areas where significant sludge l accumulation was present, and there was no sign of erosion or corrosion i damage. No loose parts or foreign objects were observed. Similar results ;

were observed at the anti-vibration bar locations. Light sludge was observed i on the top of preheater support plates (no greater than 1/16 inch thickness), !

with a thinner uniform film being present on the tubes. Preheater plate tube crevices as viewed from the top of the plates appeared unblocked in all steam generators. Some blockage was observed from the underside of the "C" tube support plates in the tube to tube support plate crevices. Limited sludge depths wer2 found on the steam generator tube sheets, with the depth of the hardened pile being on average 1/4 inch and 1/2 inch maximum. Some foreign i objects were noted by BWNT on preheater plates and on the tube sheet. _The

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most significant ubject observed was a sludge covered item, approximately 1/4 inch diameter by 4 inches long, which was lodged between tubes on the tube sheet of Steam Generator.B. The item, which was potentially_a piece of _

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welding electrode, could not be removed. During subsequent review of the eddy current examination scope (see Section 4.1), the inspectors noted that the '

tubes at the location of the lodged item had been included in the MRPC examination scope. The inspections considered this action to be reflective of -!

thorough and comprehensive engineering review during development of the eddy current examination sample. Other objects noted were: " sludge rocks"; metal shavings and chips on the 'H' preheater plate of Steam Generator D, which were ;

removed; and a ball of masking tape that was retrieved from the tube sheet of Steam Generator The inspectors considered the inspections performed by BWNT to be well-documented (BWNT Document 1226960A, Revision 0, dated June 3, 1993), with the records indicating a thorough inspection. The inspectors noted, however, that the inspection scope provided no information with respect to conditions ,

present at several of the tube support plates. Procedure OPEP07-SG-0002, Revision 0, included inspection data sheets for inspection of all tube support plates, but was not specific with respect to expectations or inspection scop . ,

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-11-3.2 Conclusions

• Procedure OPEP07-SG-00002, Revision 0, provided appropriate guidance with respect to inspection attributes to be used in visual inspection of the secondary side of steam generators, but was not specific in regard to inspection scope expectation * Contractor inspections of the Unit 2 steam ganerators appeared to have '

been well-performed for the documented inspection scop * The incorporation of tubes adjacent to a lodged foreign object in the planned MRPC eddy current examination sample was considered to be reflective of thorough and comprehensive engineering review during sample developmen '

4 REVIEW OF TUBE EXAMINATION HIFTORY, PROGRAM REQUIREMENTS, AND DATA 4.1 Review of Tube Examination History Prior to Units 1 and 2 operations, the licensee performed a full-length bobbin coil examination of all active tubes in each steam generator. During the first refueling outage (IRE 01) for Unit 1 in the Fall 1989, the licensee ]

performed a full length bobbin coil examination of a 20 percent sample of the active tubes in each steam generator. In addition, MRPC examinations were performed of all Row I U-bends. In Refueling Outage IRE 03 (Spring 1991), a full length bobbin coil examination was performed of the Technical Specifications (TS) required minimum of 6 percent in two steam generators. No repairable indications were identified during these examinations, as

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previously discussed in Section 2.4. No eddy current examinations were ;

performed during either Refueling Outage IRE 02 or 1RE0 l During the first refueling outage (2RE01) for Unit 2 in the Fall 1990, a full-length bobbin coil examination was performed cf a 4 percent sample of the active tubes in each steam generator, versus a TS minimum sample requirement of 3 percent. In Refueling Outage 2RE02 (Fall 1991), the same semple percentage was used in each steam generator. No repairable' indications were identified during these examinations, as previously discussed in Section 2.4, with one tube preventively plugged because of anti-vibration bar wear (i.e., ;

38 percent through wall wear).

The licensee adopted a comprehensive examination strategy for the current-examinations of Units 1 and 2 steam generators. Essentially, the planned examinations consisted of a full-length bobbin coil examination of all active tubes in each steam generator and a 21 percent sample for MRPC examination of l the tube expansion transition area and the ..rst and second tube support plate intersection areas. .The MRPC sample was developed to-include both randomly selected tubes and tubes with a special interest (e.g., tubes with identified fabrication anomalies, and low and high ranking based on chemical composition and mechanical properties). The inspectors considered that the sample 1

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-12-selection process used for MRPC examinations was commendable in that it '

reflected detailed engineering evaluation of all available information regarding steam generator tubing and identified fabrication anomalies. As discussed in Section 3.1, the inspectors noted that the engineering review had included the results of the visual inspection of the Unit 2 steam generators that was performed in the Spring 1993. Tubes were included, because of this review, that abutted a lodged foreign object. As noted in Section 2.4, the

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Unit 1 MRPC sample was subsequently expanded to 100 percent of the tubes at the top of the tube sheet as a result of the detection of indications in the initial 21 percent sampl ,

4.2 Review of Examination Procram Reouirements 4. Current Program The inspectors reviewed the eddy current program requirements which were contained in: (1) Engineering Instruction EI-8.01, " Steam Generator Eddy Current Data Analysis Guideline," Revision 1; Specification IUO1HWS0009,

" Specification for Eddy Current Inservice Inspection of Steam Generator Tubes," Revision 0; Engineering Instruction El-8.02, " Steam Generator Eddy-Current Data Control," Revision 1; Procedure IP-3.04Q, " Inservice Inspection Program," Revision 4; and the training manual, " Steam Generator Eddy Current Data Analyst Performance Demonstration Program," August through September 1993. The inspectors also compared the current program against prior program :

requirements and the recommendations contained in Electric Power Research Institute (EPRI) NP-6201, "PWR Steam Generator Examination Guidelines,"

Revision 2. The current overall inspection program requirements for the Units 1 and 2 steam generators werc considered good by the inspector Although no specific commitment to EPRI NP-6201, Revision 2, was noted in the .

program, it was observed that the current program and scope were, with one

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exception, generally consistent with the recommendations contained in the EPRI guidelines. The exception pertained to the absence of any program guidance concerning the EPRI NP-6201 recommendations for establishment of criteria for noisy data. Licensee personnel committed to review this subject in order to determine whether meaningful criteria could be developed to assist the data analysts. The NRC consultant noted that the primary analysis of the eddy current data was performed by BWNT, with the separate secondary analysis performed by the Zetec Computer Data Screening software. A human analyst was

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used to monitor the computer data screening. It was noted, however, that the computer data screening used a voltage threshold of 0.5 volts, although the analyst guidelines stated that no threshold was to be use *

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Criteria had been established for training and testing of data analysts that were used in the current steam generator examinations. The analysts were required to review a training manual which had been prepared for the utility by an independent consultant. This manual was considered good by the NRC consultant. One feature that was considered excellent was the referencing of defect scans (in the form of data files stored on a magneto-optical disk)

taken from actual inspections, including some from pulled tubes, rather than !

the use of only simple graphic figures for providing guidance to the analys ,

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Inclusion of some appropriately marked graphic figures in conjunction with the data files and integration of figures and tables with the text, rather than gathered at the end of the document, were considered areas where the training manual could be further improved. Both the written and practical tests for the analysts were reviewed. The written test covered the material that the analyst needed to know, but was not considered particularly challenging. The practical test was considered adequate, with a mixture of difficult calls, intermediate calls, and easy calls. There was a penalty for false calls, and if more than two repairable defects were missed, the analyst. failed the tes The inspectors ascertained that BWNT, the organization performing the eddy current examinations, had developed the practical test for the analysts. . The inspectors questioned this and were informed that this was necessitated by some of the demonstration files on a magneto-optical disk furnished by the independent consultant being found to be corrupted. The NRC consultant verified that this information was accurat ,

The NRC inspectors reviewed the testing records for the data analysts employed -

during the current outage and identified no problems with the implementation of program requirements. Hinor problems were noted with near distance visual acuity and general certification records which were referred to the licensee for resolutio . Response to Generic Communications The inspectors performed a limited review of the licensee's handling of NRC ;

generic communications pertaining to steam generator problems. The sample .;

used for this review wasBulletin 89-01, " Failure of Westinghouse Steam '

Generator Tube Mechanical Plugs," and Information Notices 90-49, " Stress Corrosion Cracking in PWR Steam Generator Tubes," and 91-67, " Problems.With the Reliable Detection of Integranular Attack (IGA) of Steam Generator Tubing." l

The review indicated that the licensee had appropriately responded to 1 Bulletin 89-01. The inspectors noted, however, with respect to Information i Notices 91-49 and 91-67 that the licensee had employed a safety risk ;

assessment type of methodology to support deferral of examinations beyond i Refueling Outage IRE 04. In part, this methodology relied on Belgian burst-test data and structural analysis, and also a prediction that only one tube with a greater than 40 percent through wall circumferential crack would occur during the fifth fuel cycle. The-inspectors did not review the bases for this rationale as a result of the current comprehensive examinations that were i being performe i

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4.2.3 Eddy Current Program Oversight The inspectors requested to see available records pertaining to licensee oversight of eddy current contractors. Only one quality assurance surveillance report, QASR 89-234, was located pertaining to a Fall 1989 ,

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-14-surveillance of Unit 1 eddy current examination activities. The. inspectors were informed by licensee personnel that engineering oversight of vendor eddy current examination activities had been performed on an ongoing basis, but not documented. Oversight was being performed of the current examinations by both Quality Assurance and Design Engineering staff, with the results documented on a formal checklis .3 Review of Tube Examination Data

The NRC consultant reviewed a sample of bobbin coil. and MRPC data that were !

obtained from the current examinations of Unit 2 steam generator tubing.~ In general, the data were observed to exhibit low noise and reflect a relatively clean state in the steam generators. Specific attention was placed on free ,

span indications that were called manufacturer's buff marks. It was noted !

during this review that software, which would permit side-by-side comparison of scans from prior outages, would be beneficial in evaluation of these types of indications. No observations were made which would bring into question

" calls" made by BWNT analyst .4 Conclusions

• With one exception, prior inservice eddy current examinations were performed using only the bobbin coil method and a sample size at or near the minimum required by the T * The licensee adopted a comprehensive examination strategy for the current steam generator examinations, performing both a full length j bobbin coil examination of all active tubes and an initial 21 percent '

MRPC sampl * The current eddy current examination program requirements were found to be good, with the primary area of improvement being the adoption of formalized training and testing of data analyst * Engineering oversight of eddy current examination contractors was undocumented for activities performed prior.to the current 1993 ;

examinations. Available records also indicated that only one quality ~

assurance surveillance had been previously performed. Engineering and-quality assurance oversight of the current examinations were being performed and documented on a checklis * The 1993 eddy current data were observed to exhibit low noise, with the ;

performance by the BWNT data analysts being found to be satisfactory for i the tube data sample that was reviewe REVIEW OF SECONDARY WATER CHEMISTRY CONTROLS AND HISTORY '!

Hany impurities thtt enter the secondary side of steam generators can contribute to corrosion of steam generator tubes and support plates. While

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-15-i the concentration of impurities needed to cause corrosion problems is normally much higher than that present in steam generator bulk water, concentration of impurities to aggressive levels is possible in occluded areas where dryout occurs. Typical areas where dryout and resulting concentration of impurities can occur are tube sheet crevices, tube support plate crevices, and sludge piles. Impurities known to contribute to tube denting (i.e., squeezing of tubes at tube supports or tube sheets as a result of the pressure of corrosion products) are chlorides, sulfates, and copper and its oxides. Pitting of steam generator tubes has been attributed to the presence of copper and concentrated chlorides. Concentrated sulfates and sodium hydroxides are believed to be major causes of intergranular SCC and IGA in steam generator ,

tubes. Iron oxide tube deposits and sludge promote local boiling and l concentration of impurities leading to these damage mechanism .1 Proaram Evolution The inspectors reviewed the licensee's secondary chemistry control program for ;

STP. Both of the STP units used the same secondary water chemistry control l program, which was documented in Procedure OPGP03-ZO-0012, " Plant Chemistry Specification," Revision 6. The plant chemistry control program incorporated :

requirements specified in the following documents:

• Updated Final Safety Analysis Report (UFSAR) Table 10.4-2, NSSS Supplier - Recommended Steam Generator Blowdown Chemistry Specifications

• UFSAR Table 10.4-6, NSSS Supplier - Recommended Condensate and Feedwater Chemistry Specifications l l

• EPRI NP-6239, "PWR Secondary Water Chemistry Guidelines," Revision 2 i dated December 1985 i

• EPRI TR-101230, " Interim PWR Secondary Water Chemistry Recommendations for IGA / SCC Control," dated September 1992 u i

The plant chemistry specifications defined normally expected ranges for key chemistry parameters, limits for critical parameters, and action levels specifying parameter devittion limits with required limitations on plant operation. The proceduro specifications used two types of parameters:

control and diagnostic. The control parameters were those parameters directly used to initiate corrective action, power reduction, or plant shutdown. The diagnostic parameters were those parameters used to provide long term trend indications and early warning of an impending chemistry control concern. .The I procedure included specifications for the steam generator water chemistry during normal operations, cold shutdown, wet layup, and prior to heat up above 200 *F. The control and diagnostic parameters were also specific to the sample location for the steam generator feedwater, blowdown, and main steam systems. The control parameters that were measured in the steam generator

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blowdown during power operations included pH, cation conductivity, sodium, chloride, sulfate, and silica. The sampling and analysis of the blowdown water chemistry for control parameters were performed on a daily basis during ,

power operation STP, Units 1 and 2, have been using hydrazine and ammonium >

hydroxide additions, respectively, for oxygen scavenging and pH control. The parameters and limits specified in the plant procedure for control of :

secondary water chemistry were found to be consistent with or more conservative than industry guidance. Historically, the licensee has incorporated both the Westinghouse and the latest EPRI recommendations for secondary water chemistry control into the STP progra .2 Secondary Side Ooeratino History .,

The inspectors reviewed the history of one steam generator for each STP unit 1 with regard to operating condition; significant chemistry, operational,.' and ,

inspection events; and compliance with the FSAR and EPRI secondary water *

chemistry guidelines. The inspectors noted that the dissolved oxygen in the feedwater was consistently within the required limit of 5 ppb or less and normally ran less than 1 pp ,

As part of this review, the inspectors obtained historical information from !

the licensee for one of the four steam generators in each unit that pertained to the water chemistry of the blowdown sampling during power operation. The i blowdown sample chemistry is summarized below in Table TABLE 1 STEAM GENERATOR BLOWDOWN CHEMISTRY PARAMETERS AT POWER OPERATION )

i UNIT 1 STEAM GENERATOR "A" ANNUAL AVERAGE CHEMISTRY VALUES ,

i PARAMETER 1992 EPRI 1989 1990 1991 1992 1993 RECOMMENDATIONS q pH > .2 9. I' .1 '

CATION 1 .2 .1 . ;

CON, uS/cm 1 (

S0DIUM, 120 5 3 4 3 10 -i ppb '

SULFATE, 520 5 1- 1 2' I ;

ppb J CHLORIDE,- $20 6 4 1 3 7 ppb SILICA, -

75 108 103 218 96 l ppb j i

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-17-UNIT 2 STEAM GENERATOR "A" ANNUAL AVERAGE CHEMISTRY VALUES PARAMETER 1992 EPRI 1989 1990 1991 1992 1993 RECOMMENDATIONS RECOMMENDATIONS pH > .1 .1 9.2 CATION CON, 1 .3 .1 0.1 uS/cm S0DIUM, ppb <20 11 4 3 2 3 SULFATE, $20 4 2 1 1 1 ppb CHLORIDE, $20 3 2 1 1 1 ppb SILICA, pab -

137 147 137 116 148 NOTE: lhe annual average is based on an average of quarterly mean values; therefore, the data may be influenced where additional sampling was performed during out-of-specification condition The historical data in Table 1 indicates that the licensee has effectively implemented the EPRI recommendations for secondary water chemistry control to ;

minimize the concentration of impurities needed to cause corrosion problems in

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steam generators since plant startup of both units. The higher than average ;

level of impurities noted in the 1993 data for Unit I were attributed by both !

the licensee and the inspectors to the sodium excursion discussed later in ;

Section '

i In addition, the information regarding the weight and chemistry of the sludge removed from each steam generator was reviewed and discussed with the licensee. The blowdown sludge data is summarized below in Table TABLE 2 STEAM GENERATOR SLUDGE ARALYSIS SG SLUDGE REMOVED, CU/FE, CHLORIDE, SULFATE, weight, lbs RATIO ppm ppm A 6 0.015 0 (wt.%) 0 (wt.%)

UNIT 1 B 6 0.015 0 (wt.%) 0 (wt.%)

1R 0 * C 6 0.015 1 (wt.%) 0 (wt.%)

D 6 0.076 0 (wt.%) 0 (wt.%)

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SG SLUDGE REMOVED, CU/FE, CHLORIDE, SULFATE, l weight, lbs RATIO ppm ppm A 1 .003 <0.12 UNIT 1 1RE03 8 21 0.002 <0.12 C 31 0.003 <0.12 D 1 .060 <0.12 A 1 .003 NDA NDA UNIT 2 B 1 .021 2 31 2RE0 9g0 C 1 .017 NDA NDA '

D 20 0.002 NDA NDA A 1 .008 .7 I UNIT 2 B 1 .007 0.39 1 RE02 gg C 1 .003 0.41 1 D .030 0.48 A 3 NDA NDA NDA UNIT 2 2RE03 B 5 NDA NDA NDA 1993 C 7 NDA NDA NDA i D NDA NDA NDA

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• 0nly X-ray fluorescence analysis for 1RE01 with compositions expressed in weight percent. Data suggests limited sensitivity and/or accurac i

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NDA - No data available on these parameter !

The small quantity of sludge removed from the steam generators and the low '

values obtained for chloride and sulfate constituents in the sludge were considered additional indications of good control of secondary water chemistry. A low copper-to-iron ratio was also noted and expected, in that, the heat exchangers in the ' secondary side do not contain copper alloys, with the exception of the aluminum bronze tube sheets in the main condense .3 Ouality Assurance and Self Assessment of the Secondary Water Chemistry l Procram I The inspectors reviewed the licensee's quality assurance audit report that was performed in April and May 1992, and the two surveillance reports that were performed during the period February through May 1993, for the secondary water

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chemistry control program. The inspectors determined that the scope of the audit and surveillances were comprehensive and appropriate for evaluation _of the implementation of the water chemistry program. The inspectors' review of the audit findings did not indicate any significant adverse findings which would bring into question the quality of the water chemistry program. The inspectors also reviewed the internal _ self assessment of the chemical operations and analysis program for trending of plant chemistry data.- The ,

assessment was performed by the primary and secondary chemists and documented in a report dated March 1993. The assessment of the trending program was comprehensive and addressed both short-term and long-term trending capabilities and performance. The report identified areas of weakness in the existing trending program regarding data management capabilities. For the interim, weekly graphing routines were subsequently set up using the current computerized program data system. For the long-term, the licensee has scheduled for 1994, the installation of an upgraded electronic data management system for on-line monitoring of Ncondary system water chemistry. The special procurement requirements for the new system included: the ability to monitor and plot real time data with alarm set points, the ability to plot short-term and long-term historical data for trending, and the capability of interfacing with other plant computers in order to integrate data such as reactor power and reactor coolant temperatur .4 Chemistry laboratory Instrumentation and In-line Process Chemistry Analyses .

The inspectors reviewed the inventory of secondary chemistry laboratory :

analytical instrumentation and in-line process chemistry analyzers installed in the various STP water systems. The inspectors verified that the necessary instrumentation was installed in the process line or available in the laboratory for the analysis of the diagnostic and control parameters specified in the secondary water chemistry control program. Several in-line process and laboratory it.struments had been upgraded to state-of-the-art analytical !

instrumentation to perform the required chemical analyses. The original in-line chemistry monitoring capability was upgraded: by the addition of an . .

in-line ion chromatograph in May 1990; by the replacement of the blowdown and feedwater in-line cation conductivity analyzers in 1990 and 1992, >

respectively; by the replacement of the in-line dissolved oxygen analyzer on the condensate polisher discharge line in January 1992; and by the replacement of all in-line silica analyzers in February 1993. All of these analyzer upgrades and additions were efforts made by the licensee to enhance their ability to monitor secondary water chemistry conditions which could affect th :

steam generator .5 Off-Normal Secondary Chemistry History The inspectors reviewed documentation of out-of-specification secondary water chemistry conditions for the period 1989 through first quarter 1993.- Two '

significant out-of-specification chemistry conditions were noted which could contribute to the degradation of steam generators. They were both out-of- '

specification sodium conditions in the Unit I secondary water system. The ,

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first condition occurred in 1990 after a Unit I reactor trip. When the auxiliary feedwater system was used for transition to " wet layup" of the steam generators, sodium hydroxide was inadvertently introduced to the auxiliar feedwater storage tank from a drum labeled hydrazine. Steam generator sodium concentrations ranged from 169 ppb to 3900 ppb. Corrective actions were taken and the sodium concentrations were reduced to within the specified control limit of 100 ppb within 2 days. The second out-of-specification' sodium ,

condition occurred in 1993 when Unit I was in MODE I at 48% power increasing '

to 100% power. Sodium and chlorine entered the system from " poor quality" water which accumulated in the flash tank of the Low Pressure Heater Drain Pump No. 12 that was pumped into the steam generator. feedwater when a valve did not stroke fully closed. Sodium and chloride concentrations in the steam generator blowdown sampling increased to 180 ppb and 120 ppb, respectivel The steam generator blowdown flow was maximized and sodium and chloride concentrations were within the specified control limit of 20 ppb in approximately 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .6 Conclusions ,

• Since commercial operation of STP, Units 1 and 2, the secondary water chemistry program for both units had continually been upgraded to incorporate industry guidelines as they were made availabl * Several in-line process and laboratory instruments had been upgraded to state-of-the-art analytical instrumentation to enhance the performance of required chemical analyses. These upgrades included the addition of an ion chromatograph and replacement of analyzers for cation conductivity, dissolved oxygen, and silic * The licensee has maintained excellent control of the secondary water chemistry, with only two significant out-of-specification chemistry conditions noted since plant startup. These conditions both involved out-of-specification-sodium concentrations that occurred in Unit I during 1990 and again in 1993. In each case, the out-of-specification condition was promptly identified and corrected before required escalation to Action Level >

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ATTACHMENT l

1 PERSONS CONTACTED {

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1.1 Licensee Personnel _

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• R. Baker, Senior Consulting Engineer, Design Engineering .

• H. Bergendahl, Manager, Technical Services  ;

• Blair, Manager, Staff Training Division

• D. Bryant, Chemical Analysis General Supervisor l

• Chakravorty, Executive Director, Nuclear Safety Review Board ;

• T. Cloninger, Vice President, Nuclear Engineering  !

• R. Cook, Industry Relations  ;

• W. Cottle, Group Vice President

• • M. Coughlin, Senior Licensing Engineer i

• M. Ebel,Section XI Coordinator '

• • J. Fox, Design Engineer -

G. Gibson, Staff Nuclear Chemist

• J. Haning, Staff Engineer, Design Engineering i

• W. Harris,Section XI Supervisor  :

• S. Hubbard, Quality Control Supervisor i

• J. Johnson, Supervisor, Quality Assurance

• T. Jordan, General Manger, Nuclear Engineering l

• D. Keating, Director, Independent Safety Engineering Group t R. Kersey, Design Engineer -

• D. Leazar, Manager, Plant Engineering  :

• L. Martin, General Manager, Nuclear Assurance

• G. Parkey, Plant Manager _

• P. Parrish, Senior Licensing Specialist  ;

• • U. Patil, Supervising Engineer, Design Engineering '

• s. Sheppard, General Manager, Nuclear Licensing  ;

• B. Tran, Check Valve Program Coordinator -i

• C. Walker, Manager, Public Information -

• P. Walker, Staff Engineer '

• J. Wigginton, Associate Engineer, Licensing -

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1.2 NRC Personnel i L

• D. Loveless, Senior Resident Inspector

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• T. Westerman, Section Chief, Division of Reactor Safety '

i In addition to the personrel listed above, the inspectors contacted other' I personnel during this inspectio * Denotes personnel attending the August 27, 1993,. exit meetin ** Denotes personnel attending both the August 27 and October 8, 1993, exit -

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s-2-2 EXIT MEETING An initial exit meeting was conducted on August 27, 1993. During this meeting, the inspectors reviewed the scope and findings'of.the report with respect to the eddy current program and implementation for Unit 2. Two documents were reviewed during the inspection which were identified by the licensee as being considered proprietary by its vendor. No information was, included in the inspection report that was considered proprietary. - A fina exit meeting was held telephonically on October 8,1993, with respect _ to Unit I eddy current result h W

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