05000321/LER-2010-001, Regarding Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift

Regarding Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift

ML101090069
Person / Time
Site:	Hatch
Issue date:	04/16/2010
From:	Madison D Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-10-0760 LER 10-001-00
Download: ML101090069 (6)

LER-2010-001, Regarding Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function
3212010001R00 - NRC Website

text

Dennis R. Madison Southern Nuclear Vice Presidenl* Hatch Operating Company, Inc.

Plant Edwin! Hatch 11028 Halch fOarkway North Baxley. Georgia 31513 T(;1912.537.5859

!"ax 912.366.2077 SOUTHERNA COMPANY April 16,2010 Docket No.:

50-321 NL-10-0760 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), Southern l\\Juclear Operating Company is submitting the enclosed Licensee Event Report (LER) concerning safety relief valves allowable test range exceeded due to corrosion induced bonding.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, p~~vr;~-./

D. R. Madison Vice President - Hatch DRM/M..IKlmsc Enclosure: LER 1-2010-001 cc:

Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. E. D. Morris, Senior Resident Inspector - Hatch

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 9-2007)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

13. PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000321 1 OF 5

14. TITLE Corrosion Induced Bonding Results in Safety Relief Valve Lift Setpoint Drift

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 03 08 2010 2010 - 001 -

0 04 16 2010 05000

~. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 5 o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) o 20.2203(a)( 1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)

10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)( 1)(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71 (a)(4) 000 o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71 (a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi)

I8J 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in NRC Form 366A

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME I~ELEPHONE NUMBER (Include Area Code)

IEdwin I. Hatch / Steve Tipps, Principle Licensing Engineer 912-537-5880 CAUSE SYSTEM COMPONENT MANU REPORTABLE

CAUSE

SYSTEM COMPONENT MANU REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B

SB RV T020 Yes

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED SUBMISSION MONTH DAY YEAR o YES (If yes, complete 15. EXPECTED SUBMISSION DATE)

I8J NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On March 11, at approximately 1300 EST, Unit 1 was at 0 CMWTh, which is 0.00 percent of rated thermal power (RTP). On that day it was concluded that at the completion of bench testing five Safety Relief Valves (SRVs) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit.

The root cause of the SRV setpoint drift is corrosion-induced bonding between the pilot disc and seating surface.

Immediate corrective actions for this event included replacement of all eleven SRV pilot valves with refurbished pilot valves which have pilot discs made from Stellite 21 material. Also improvements in the insulation surrounding each SRV were made to reduce the likelihood of corrosion-induced bonding between the pilot disc and seating surface.

PRINTED ON RECYCLED PAPER NRC FORM 366 (9-2007)

(If more space is required, use additional copies of NRC Form 366A)

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).

DESCRIPTION OF EVENT

On March 11, at approximately 1300 EST, Unit 1 was at 0 CMWTh, which is 0.00 percent of rated thermal power (RTP). On that day it was concluded that at the completion of bench testing five Safety Relief Valves (SRVs, EllS Code SB) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit which is +/- 3%. The setpoint for each of the eleven SRVs is 1150 +/- 34.5 psig. The following is a tabulation of the test results for the eleven SRVs:

MPLNumber 1B21-F013A 1B21-F013B 1B21-F013C 1B21-F013D 1B21-F013E 1B21-F013F 1B21-F013G 1B21-F013H 1B21-F013J 1B21-F013K 1B21-F013L Pilot Serial Number 313 1006 306 304 1007 303 1228 310 1231 1004 1189 As-Found Lift Pressure 1147 1167 1250 1165 1205 1168 1239 1160 1297 1208 1173 Percent Drift 99.7 101.5 108.7 101.3 104.8 101.6 107.7 100.9 112.8 105.0 102.0 These valves were removed from service during the Spring 2010 refueling outage and replaced with like kind valves that were serviced and tested in accordance with plant procedures.

PRINTED ON RECYCLED PAPER (9-2007)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE Edwin 1. Hatch Nuclear Plant Unit 1 05000321 YEAR I

SEQUENTIAL IREVISION NUMBER NUMBER 3

OF 5

2010 001 0

CAUSE OF EVENT

The cause of the SRV setpoint drift exceeding the allowable plant TS limit is corrosion induced bonding between the pilot disc and seating surface. Proper insulation of the SRV's has an impact on the formation of corrosion on the pilot seats. This impact is addressed in GE SIL 169, Supplement 16. Proper insulation improves the environment in the pilot area.

Per the GE SIL, higher temperature in the pilot area results in higher concentration of steam in that area, which further results in lower concentration of oxygen. The lower concentration of oxygen reduces the amount of corrosion that will occur in the seating area. The as-found condition of the Unit 1 insulation revealed that the insulation was in poor condition, that only one layer of blankets was used, and that the blankets were sagging down around the pilot flange.

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 50.73(a)(2)(i)(B) because an event occurred which is prohibited by Technical Specifications (TS). Specifically, multiple test failures of the SRVs is defined as reportable in NUREG-1022, Revision 2, dated October 2000, in section 3.2.2, example 3, titled "Multiple Test Failures."

The 11 SRVs, which are located on the four main steam lines within the drywell (EllS Code NH) between the reactor vessel (EllS Code AD) and the inboard main steam isolation valves (MSIV EllS Code SB), are required during Modes 1,2, and 3 to limit the peak pressure in the nuclear system such that it will not exceed the applicable ASME Boiler and Pressure Vessel Code limits for the reactor coolant pressure boundary. Per TS Surveillance Requirement 3.4.3.1, the valves are tested in accordance with the In-service Testing Program to verify the safety function lift setpoints are within the specified limits. The SRVs must accommodate the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code limit of 1375 psig peak vessel pressure, has been defined as a closure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV position switches; the reactor ultimately shutdowns from a high neutron flux trip. Analysis of this event comparing the as-found bench test results for SRV actuation pressures with previously analyzed lifting pressures for each SRV has demonstrated that the resultant peak pressure was within the ASME Code limit of 1375 psig. The previously analyzed lifting pressures were evaluated and determined to be bounding in comparison to the as-found lift points for this event. Furthermore, the plant TS overpressure safety limit of 1325 psig dome pressure must be met during normal operations and for anticipated operational occurrences (ADOs). The analysis of the as-found test results by comparison with previously analyzed lifting pressures also showed that the resultant dome pressure was within the plant TS Safety Limit.

PRINTED ON RECYCLED PAPER u.s. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

(~2007)

CONTINUATION SHEET

3. PAGE

2. DOCKET

6. LER NUMBER

1. FACILITY NAME SEQUENTIAL I REVISION YEAR I

NUMBER NUMBER 4

OF 5

Edwin I. Hatch Nuclear Plant Unit 1 05000321 2010 001 o

In addition, a non-credited electrical actuation system was installed in 1993 to ensure proper actuation of the SRVs. This system provides a redundant, independent method (i.e., electrical signal) to actuate the SRVs. During the run cycle the redundant electrical system was available. The system was procured to Class IE environmental and seismic standards, and is deemed highly reliable.

Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety.

CORRECTIVE ACTIONS

All eleven pilot valves have been replaced with refurbished pilot valves.

Each of the eleven pilot disc from the valves removed were replaced with a pilot disc made from Stellite 21 material.

Insulation surrounding each SRV has been upgraded to improve resistance to corrosion induced bonding.

ADDITIONAL INFORMATION

Other Systems Affected: None

Failed Components Information

Master Parts List Number: lB2l-F0l3 EllS System Code: SB Manufacturer: Target Rock Reportable to EPIX: Yes Model Number: 7567F Root Cause Code: B Type: Relief Valve EllS Component Code: RV Manufacturer Code: T020 Commitment Information: This report does not create any new permanent licensing

Previous Similar Events

LER 2-2009-001, identified multiple SRV setpoint drift for five of the eleven SRV's.

Corrective actions for this event replaced all eleven pilot discs with Stellite 21 material, and improvements were made in the insulation surrounding the SRV's. These actions were taken on Unit 2 therefore they do not directly impact Unit 1 performance. However improvement in the insulation was incorporated into corrective action for the current LER, 1-2010-001.

PRINTED ON RECYCLED PAPER (9-2007)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000321 YEAR I

SEQUENTIAL I REVISION NUMBER NUMBER 5

OF 5

2010 001 0

LER 2-2008-004; identified multiple SRV setpoint drift for three of the four tested SRV's.

Four SRV's were removed mid-cycle. Corrective actions for this LER, replacement of discs were implemented but discs made of stellite 21 for the Unit 2 SRV's were not available for all of the replaced discs. These actions were taken on Unit 2 therefore they do not directly impact Unit 1 performance.

LER 1-2008-002; identified multiple SRV setpoint drift for three of the eleven SRV's.

Corrective action for this LER, replacement of discs with stellite 21 discs, was completed during the 2008 Unit 1 refueling outage. Industry experience indicates the stellite 21 material to be resistive to corrosion induced bonding. The initial experience for plant Hatch did not follow the industry experience. During the 2008 refuel outage the condition of insulation surrounding the SRV's was not closely controlled.

LER 2-2007-006, identified multiple SRV setpoint drift for five of the eleven SRV's.

Corrective action for this LER was replacement of discs with stellite 21 discs. The discs were replaced during the 2009 Unit 2 refueling outage. These actions were taken on Unit 2 therefore they do not directly impact Unit 1 performance.

LER 1-2006-003, which identified an error in reporting multiple SRV setpoint drift, also described results from the previous three outages where multiple SRV setpoint drift had occurred. Corrective actions for that LER focused on ensuring proper reporting of SRV setpoint drift was performed.

PRINTED ON RECYCLED PAPER