05000275/LER-2005-001

LER-2005-001,

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)
2752005001R00 - NRC Website

EXPECTED

SUBMISSION DATE (15) 1 1 1 MON - DAY YR On November 11, 2005, at 1200 PST, with Unit 1 in Mode 6 (refueling, with the reactor defueled), analysis of eddy current testing on Steam Generator (SG) 1-1 and SG 1-2 indicated that greater than one percent of the tubes were defective.

Pacific Gas & Electric Company (PG&E) made a non-emergency report to the NRC as required by Technical Specification (TS) Table 5.5.9-2, "Steam Generator (SG) Tube Inspection," at 1230 PST, via NRC Event No. 42136.

On November 9, 2005, PG&E discussed the preliminary results of the SG eddy current inspection during a phone conference with the Office of Nuclear Reactor Regulation (NRR). The primary cause of the SG tube cracking was circumferential outside diameter stress corrosion cracking at the hot leg tube support plates and at the hot leg top of tubesheet.

In accordance with TS 5.5.9, "Steam Generator (SG) Tube Surveillance Program," PG&E has plugged all defective Unit 1 tubes identified during the current refueling outage. All defective tubes met condition-monitoring� .

requirements at the end of Cycle 13. PG&E maintains a comprehensive program to minimize SG tube degradation, and plans to replace the SGs at the end of Cycle 15.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6)

I. Plant Conditions

Unit 1 was in Mode 6 (refueling, with the reactor defueled) in its thirteenth refueling outage (1R13).

A. Background

Technical Specification (TS) 5.5.9, "Steam Generator (SG) Tube Surveillance Program," requires that the results of each SG tube inspection be classified as Category C-3 if more than one percent of the total tubes inspected are defective. Defective tubes have service induced degradation in excess of TS tube repair limits. Defective tubes must be removed from service by plugging.

TS 5.6.10, "Steam Generator (SG) Tube Inspection Report," paragraph a, requires the number of tubes plugged in each SG [AB][TBG] to be reported within 15 days following the completion of each inservice inspection. TS 5.6.10, paragraph c, requires the results of SG tube inspections, which fall into Category C-3, to be reported in a special report to the Commission within 30 days and prior to resumption of plant operation.

If the results of the SG tube inspections are classified as Category C-3, then NRC notification is required in accordance with TS Table 5.5.9-2, "Steam Generator (SG) Tube Inspection," and submittal of a special report is required in accordance with TS 5.6.10.c.

B. Event Description

On November 9, 2005, PG&E discussed the preliminary results of the inspection during a scheduled phone conference with the NRR.

On November 11, 2005, at 1200 PST, final analysis of eddy current testing on SG 1-1 and SG 1-2 indicated that greater than one percent of the tubes inspected were defective, therefore classifying the SGs as Category C-3, per TS 5.5.9. SG 1-1 had 37 defective tubes and SG 1-2 had 41 defective tubes that were removed from service by tube plugging during 1R13.

Additional degraded tubes were also preventively plugged.

On November 11, 2005, a non-emergency report was made in accordance with TS Table 5.5.9-2 and 10 CFR 50.72(b)(3)(ii)(A), at 1230 PST.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) C. Status of Inoperable Structures, Systems, or Components that Contributed to the Event None.

D. Other Systems or Secondary Functions Affected None.

E. Method of Discovery The defective tubes were found during routine scheduled eddy current testing of Unit 1 SG tubing performed during 1R13.

F. Operator Actions None.

G. Safety System Responses None.

III.�Cause of the Problem A. The number of defective tubes in SG 1-1 and SG 1-2 identified during 1R13 exceeded one percent of the total tubes inspected, placing the SGs in Category C-3, per TS 5.5.9.

B. Root Cause The majority of the tube defects are attributed to axial outside diameter stress corrosion cracking (ODSCC) at the hot leg tube support plate, and circumferential ODSCC at hot leg tube support plate and at the hot leg top of tubesheet. In addition, axial ODSCC at the hot leg top of tubesheet was identified at Diablo Canyon Power Plant (DCPP) for the first time.

C. Contributory Cause None.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) OF I 6Diablo Canyon Unit 1 0 I5 0 0 0 5 2005 0 0 1 0 0 42 I 7

IV. Assessment of Safety Consequences

The licensing-basis, large-break, loss-of-coolant accident analysis assumes a tube-plugging limit of 15 percent per SG. Including the tubes plugged during 1R13, the following table presents the number of tubes (out of a total of 3,388 tubes for each SG) that are currently plugged in each of the four Unit 1 SGs.

SG Tubes Plugged in Total Tubes Total Percentage of 1-1 38 232 6.8 1-2 43 315 9.3 1-3 16 92 2.7 1-4 19 186 5.5 The plugging percentage for each Unit 1 SG remains within the current allowable limit of 15 percent.

There were no actual safety consequences involved in this event since all defective tubes met the performance criteria of the Nuclear Energy Institute (NEI) 97-06, Revision 2, "Steam Generator Program Guidelines.

Also, the condition is not considered a Safety System Functional Failure.

Therefore, the event is not considered risk significant and it did not adversely affect the health and safety of the public.

V. Corrective Actions

A. Immediate Corrective Actions

The Unit 1 SG tubes classified as defective during 1R13 have been plugged in accordance with DCPP TS 5.5.9.

B. Corrective Actions to Prevent Recurrence

PG&E has initiated several programs to minimize SG tube degradation and implemented alternate repair criteria to leave certain indications in service, thus extending the SG service life. Circumferential ODSCC indications are not subject to alternate repair criteria (ARC), and must be plugged upon detection.

PG&E maintains a comprehensive program to minimize SG tube degradation, and plans to replace the SGs at the end of Cycle 15.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION � FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) OFDiablo Canyon Unit 1 0 5 0 0 0 1 2 7 5 2005 0 l 0 0 5 6 Secondary side initiatives to minimize tube degradation:

1. EPRI secondary chemistry recommendations were implemented to minimize ODSCC at tube support plates (TSPs) (e.g., hydrazine levels were increased in 1992, the secondary side pH treatment was converted from ammonia to ethanol amine, and a molar ratio control program was implemented in 1993/1994).

2. Tube sheets have been sludge lanced during each refueling outage to minimize ODSCC at the tube sheet.

3. DCPP has an upgraded plant makeup water system to minimize SG contaminate levels.

4. SG blowdown is maintained at approximately one percent of the main steaming rate to minimize SG contaminate levels.

5. A boric acid addition program is in effect, including boric acid soaks at startup to mitigate denting and ODSCC at TSPs.

6. DCPP has condensate polishers and emergency (plant curtailment) procedures to protect against seawater condenser tube leaks.

7. The Unit 1 SGs were chemically cleaned in 2004.

Primary side initiatives to minimize tube degradation:

1. Rows 1 and 2 U-bends were heat treated in 1987/1988 to prevent primary water stress corrosion cracking (PWSCC).

2. The tubes in the hot leg tube sheet region were shot peened in 1992/1993 to minimize PWSCC.

3. Reactor coolant system (RCS) contaminants are maintained at low levels- in accordance with EPRI guidelines.

4. Lithium and boron concentrations are coordinated to minimize pH swings in the RCS.

5. Zinc addition to the RCS was implemented in Units 1 and 2 starting in Cycle 9 to inhibit PWSCC in SG tubes.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Diablo Canyon Unit 1 0 151 0 0 1 0 7 5 2005 0 0 i 1 0 0 6 I OF 6 Alternate repair criteria (ARC):

1. Voltage-based ARC for axial ODSCC at TSPs was implemented starting in the Unit 2 eighth refueling outage and the Unit 1 ninth refueling outage (1R9).

2. W* ARC for axial PWSCC contained in the WEXTEX tube sheet was implemented starting in 1R9 and the Unit 2 ninth refueling outage.

3. PWSCC ARC for axial PWSCC at dented TSPs was implemented starting in the Unit 1 eleventh refueling outage (1R11) and the Unit 2 eleventh refueling outage.

VI.WAdditional Information

A. Failed Components

Component:WSG tubes (Series 51 SG) Manufacturer: Westinghouse

B. Previous Similar Events

inspected in SG 1-1 and 1-2, during the Unit 1 eighth refueling outage, were defective.

inspected in SG 1-2, during the Unit 1 tenth refueling outage, were defective.

inspected in SG 1-2, during 1R11, were defective.

inspected in SG 1-1 and SG 1-4, during Unit 1 twelfth refueling outage, were defective.