05000321/LER-2014-003, Regarding Safety Relief Valves as Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

Regarding Safety Relief Valves as Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

ML14188C020
Person / Time
Site:	Hatch
Issue date:	07/07/2014
From:	Vineyard D Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-14-0996 LER 14-003-00
Download: ML14188C020 (6)

LER-2014-003, Regarding Safety Relief Valves as Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(x) 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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3212014003R00 - NRC Website

text

David R. Vineyard Vice President

• Hatr:h July 7. 2014 Southern Nuclear Operating Company, Inc.

Pl;~nt Edwin I. llatch 11 02811atch Parkway North Baxley, Georgia 31513 Tel912.537.5859 Fax 912.366.2077 Docket Nos.: 50-321 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report 2014-003-00 SOUTHERN A COMPANY NL-14-0996 Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria Ladies and Gentlemen:

In accordance with the requirements of 10 CFR 50. 73(a)(2)(i)(B) Southern Nuclear Operating Company hereby submits the enclosed Licensee Event Report.

This letter contains no NRC commitments. It you have any questions, please contact Greg Johnson at (912) 537-5874.

Respectfully submitted, D. R. Vineyard Vice President - Hatch DRV/mr Enclosures: LER 2014-003-00

/

U. S. Nuclear Regulatory Commission NL-14-0996 Page2 cc:

Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President-Hatch Mr. B. L. Ivey, Vice President-Regulatory Affairs Mr. D. R. Madison, Vice President-Fleet Operations Mr. B. J. Adams, Vice President-Engineering Mr. G. L. Johnson, Regulatory Affairs Manager-Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatorv Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager-Hatch Mr. David Hardage, Senior Resident Inspector-Hatch

NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150.0104 EXPIRES: 01/3112017 (02*2014)

Es5mated burden par response to comply y,jjh this manda!OI}' cdlecticn request 80 tlours *

\\_~) UCENSEE EVENT REPORT (LER)

Reported lessens learned a~e incclpcnlted into the licensing process and fed bad< to industly.

Send commenls regarding burden estimate to the FOIA, Privacy and lnfonnaticn Co!lectiOnS Blanch (T-5 F53), U.S. Nucle81 Regula!QIY Ccmmission. Washington, DC 20555-0001, Clf by internet e-mail to lnlocol!eds.ResourceOnn:.gov, and to file Desk Officer, Office ollnformalion and (See Page 2 for required number of RegUalcly Al!airs. NE0&10202. (3150-0104~ Ollice al Managemantand Budget. Waslling1cn, DC digits/characters for each block) 20503. If a means used to impose an inlonnalion cdlecticn does not dsJ:lay a curren!ly valid OMB ccn1ro1 runber, file NRC may not CCII'Gicl or sponsor, and a person is net required to respond to.

the information cdleclion.

3.PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000 321 1 OF 4

4. TITLE Safety Relief Valves As Found Settings Resulted In Not Meeting Tech Spec Surveillance Criteria

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

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

MONTH DAY YEAR FACIUTY NAME DOCKET NIJMBER 05 07 2014 2014 - 003 - 00 7

7 2014

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply}

D 20.2201

0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii)

Mode 1 D 2o.22o1(d) 0 20.2203(a)(3)(il) 0 50.73(a)(2)(11)(A) 0 50. 73(a)(2)(viii)(A) 0 20.2203(a)(1) 0 20.2203(a)(4)

D 50. 73(a)(2)(1i)(B) 0 50.73(a)(2)(viii)(B) 0 20.2203(a)(2)(1) 0 50.36(c)(1)(i)(A) 0 50. 73(a)(2)(11i) 0 50.73(a)(2)(1x)(A)

10. POWER LEVEL 0 20.2203(a)(2)(il) 0 50.36(c)(1)(ii)(A) 0 50. 73(a)(2)(iv)(A) 0 50.73(a)(2)(x) 0 20.2203(a)(2)(iil) 0 50.36(c)(2) 0 50. 73(a)(2)(v)(A) 0 73.71(a)(4) 100 0 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) 0 73.71(a)(5) 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C)

D OTHER 0 20.2203(a)(2)(vi) 181 50.73(a)(2}(i)(B) 0 50.73(a)(2)(v)(D)

Specify in Absltad below Clf In NRC Fonn 366A

12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT

~~LEPHONE NUMBER (lneluds Alea CodB)

Edwin I. Hatch I Steven Tipps - Licensing Supervisor 912-537-5880 CAUSE SYSTEM COMPONENT MANU*

REPORTABLE

CAUSE

SYSTEM COMPONENT MANU*

REPORTABLE FACTURER TOEPIX FACl\\JRER TOEPIX B

SB RV T020 y

~:

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH DAY YEAR 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 181No SUBMISSION DATE

~STRACT (Umit to 1400 spaces, I.e ** approximately 15 s/ngfe*spaced type.winen tines)

On May 7, 2014, at approximately 0837, Unit 1 was at 99.9 percent rated thermal power (RTP) when the "as-found" testing results of the 2-stage main steam safety relief valves (SRVs) were received which indicated that five of eleven SRVs had experienced setpoint drift during the previous operating cycle which resulted in their failing to meet the Technical Specification (TS) opening setpoints of 1150 psig +1-3 percent as required by TS surveillance requirement 3.4.3.1.

The root cause of the SRV setpoint drift is attributed to corrosion-induced bonding between the pilot disc and seating surfaces. This conclusion is based on previous root cause analyses and the repetitive nature of this condition at Hatch and within the BWR industry. The 2-stage SRVs with platinum coated pilot seats were removed from Unit 1 during the 2014 refueling outage and replaced with 3-stage SRVs with a modified pilot. 3-stage SRVs typically do not exhibit set point drift, additionally the modified pilot reduces instances of vibration induced spurious openings and leak-by.

A 3-stage SRV with a similar modified pilot was installed on Unit 2 during the 2013 outage. Current plans are to replace the remaining ten valves at Unit 2 with the same modified pilot valves during the next outage in 2015.

NRC FORM :Mi6 (02*2014)

~RC =

PLANT AND SYSTEM IDENTIFICATION

General Electric

• Boiling Water Reactor

6. LER NUMBER I

SEQUENTIAL I NUMBER REV NO.

2014 003 00 Energy Industry Identification System codes appear in the text as (EllS Code XX).

DESCRIPTION OF EVENT

3.PAGE 2

OF On May 7, 2014, at approximately 0837, Unit 1 was at 99.9 percent rated thermal power (RTP) when the report of the *as-found" testing results of the 2-stage main steam safety relief valves (SRVs) were received which Indicated that five of eleven SRVs (EllS Code SB) had experienced setpoint drift during the previous operating cycle which resulted in their allowable TS surveillance requirement (SA) 3.4.3.1 limits of 1150 +1-34.5 psig (+/* 3 percent) being exceeded. The following Is a tabulation of the test results of the eleven SRVs:

MPLNumber Pilot Serial Number As-Found Lift Pressure Percent Drift (psia) 1821*F013A 1007 1175 2.17 1B21*F013B 1010 1172 1.91 1821*F013C 1003 1223 6.35 1821*F013D 311 1197 4.09 1B21-F013E 1226 1181 2.70 1B21*F013F 1011 1168 1.57 1B21*F013G 314 1181 2.70 1821*F013H 312 1166 1.39 1821*F013J 304 1204 4.70 1B21*F013K 301 1198 4.17 1B21*F013L 306 1201 4.43 All eleven valves were removed from service during the spring 2014 refueling outage and replaced with 3-stage SRVs. The 3-stage SRVs have a modified pilot that helps reduce the possibility of inadvertent lift and leak by.

4 The 2-stage SRVs installed on Unit 2 prior to the 2013 refueling outage utilized platinum coated pilot discs. Even though these valves are identical to the valves that were used in Unit 1 Cycle 26, there is confidence that the 2-stage SRVs currently installed on Unit 2 will function reliably for the remainder of the cycle. This Is based on the recent Unit 2 operating experience from the previous outage (10 of 11 SRVs met Tech Spec acceptance criteria), the fact that the SRVs were successfully tested prior to installation and that new platinum coated pilots were installed. Plans are to replace the existing 2-stage SRVs on Unit 2 with 3-stage SRVs during the next scheduled refueling outage as a long term corrective action.

CAUSE OF EVENT

The root cause of the SRV setpoint drift is attributed to corrosion-induced bonding between the pilot disc and its seating surface. This conclusion is based on previous root cause analyses and the repetitive nature of this condition at Plant Hatch and in the industry. In General Electric (GE) service information letter (SIL) 196, Supplement 16, GE determined that condensation of steam in the pilot chamber of Target Rock 2-stage SAVs can cause oxygen and NRC FOAM 3GIA 102*2014)

!NRC FORM 366A 02*20141 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150.0104 f~*)

~---*~ LICENSEE EVENT REPORT (LEA)

CONnNUATION SHEET

1. FACIUTY NAME 2.DOCKET Edwin I. Hatch Nuclear Plant Unit 1 05000321 YEAR 2014

8. LER NUMBER I

SEQUENTIAL I NUMBER 003 REV NO.

00 EXPIRES: 011.1112017 3.PAGE 3

OF 4

hydrogen dissolved in the steam to accumulate. As steam condenses in the relatively stagnant pilot chamber, the dissolved gases are released. In a volume such as the pilot chamber which is normally at approximately a 1000 psig pressure and a temperature of 545 degrees F, the total pressure consists primarily of water vapor partial pressure because 544.6 degrees F is the saturation temperature at 1000 psig. This wet, hot, high-oxygen atmosphere can be very corrosive and can increase the likelihood of corrosion-induced bonding of the pilot disk to its seat. It was also noted that proper insulation minimizes the accumulation rate of non-condensable gases and the steady-state oxygen partial pressure. Despite improvements made in maintaining the integrity of insulation for the previously installed 2-stage SRVs the corrosion-induced bonding continued to occur as evidenced by the test results from this most recent outage.

REPOATABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable in accordance with Title 10 of the Code of Federal Regulations (CFR), Part 50.73(a)(2)(i)(B) because an event occurred which Is prohibited by TS surveillance requirement (SA) 3.4.3.1. Specifically, an example of multiple test failures is given in NUREG-1022, Revision 3, "Event Reporting Guidelines 10 CFR 50.72 and 50.73" which describes the sequential testing of safety valves. This example notes that "Sometimes multiple valves are found to lift with set points outside of technical specification limits."

NUREG-1022 further states in the example that "discrepancies found in TS surveillance tests should be assumed to occur at the time of the test unless there is firm evidence, based on a review of relevant information (e.g., the equipment history and the cause of failure), to indicate that the discrepancy occurred earlier. However, the existence of similar discrepancies in multiple valves is an indication that the discrepancies may well have arisen over a period of time and the failure mode should be evaluated to make this determination." Based on this guidance and the fact that the development of the corrosion occurred over a period of time of plant operation, the determination was made that this "as fo.und" condition is reportable under the reporting requirements of 10 CFA 50.73(a)(2)(i)(B).

There are eleven SRVs located on the four main steam lines within the drywall (EllS Code NH) in between the reactor pressure vessel (EllS Code AD) and the inboard main steam isolation valves (MSIV, EllS Code SB). These SRVs are required to be operable 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. The SAVs are tested in accordance with TS surveillance requirement 3.4.3.1 in which the valves are tested as directed by the In-Service Testing Program to verify lift set points are within their specified limits to confirm they would perform their required safety function of overpressure protection. 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 by an event involving the closure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV position switches with the reactor ultimately shutting down as the result of a high neutron flux trip (a scenario designated as MSIVF).

This MSIVF event analysis was performed for the Unit 1 Cycle 24 "as-found" condition of the SRVs showed that there was adequate margin to the vessel dome pressure and ASME vessel overpressure limits. Another analysis was performed which compared the Cycle 26 "as-found" SRV measured opening pressures to those of the Cycle 24 reload licensing analysis (ALA). The results from this analysis showed a decrease in peak pressures due to the fact that eight of the eleven SAVs opened at pressures lower than those which were measured in the Cycle 24 RLA.

Based on this comparison, recognizing that SRV opening pressures used In the Cycle 24 ALA were generally much larger than the Cycle 26 measured pressures, and since the Cycle 24 ALA showed sufficient margin to the Technical Specification 2.1.2 dome pressure safety limit and the ASME Pressure Vessel Code overpressure limit, the results of the Cycle 24 remain bounding. Therefore the peak pressure at the bottom of the vessel remained below the ASME (02*20141 U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

1. FACILITY NAME 2.DOCKET

6. LER NUMBER 3.PAGE OF YEAR I SEQUENTIAL I REV NUMBER NO.

4 Edwin I. Hatch Nuclear Plant Unit 1 05000 321 2014 003 00 Boiler and Pressure Vessel code limit, and the peak reactor pressure vessel dome pressure remained within the Tech Spec Safety Limits.

Additionally, a highly reliable, though non-credited, electrical actuation system serves as a redundant, independent method to actuate the SAVs. During Cycle 26 this redundant electrical logic system was fully functional.

4 Based on the analyses performed, the overpressure protection system would have continued to perform its required safety function if called upon in its "as found" condition. Therefore, this event had no adverse Impact on nuclear safety and was of very low safety significance.

CORRECTIVE ACTIONS

The 2-stage SRVs with platinum-coated pilot discs were removed from Unit 1 during the 2014 refueling outage and replaced with 3-stage SRVs that have a modified pilot. 3-stage SRVs typically do not exhibit set point drift due to their design. The modified pilots will help reduce spurious openings and leak-by due to system vibration.

Additionally, current plans are to replace the remaining 2-stage Unit 2 SRVs with 3-stage SAVs with the same modified pilots during the 2015 refueling outage.

ADDITIONAL INFORMATION

Other Systems Affected: None

Failed Components Information

Master Parts List Number: 1 B21*F013C, D. J, K, L Manufacturer: Target Rock Model Number: 7567F Type: Relief Valve Manufacturer-Code: T020 EllS System Code: SB Reportable to EPIX: Yes Root Cause Code: B EllS Component Code: RV Commitment Information: This report does not create any licensing commitments.

PREVIOUS SIMILAR EVENTS

LEA 1-2012*004, identified multiple SRV setpoint drift for 8 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 2-stage SRVs whose pilot discs had undergone a platinum surface treatment which was considered at that time to be the long term fix for this corrosion bonding issue.

LEA 2-2011-002, identified multiple SRV setpoint drift for 8 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs during the Unit 2 Spring 2011 refueling outage which was considered at that time to be the long term fix for this corrosion bonding issue. Subsequent to that outage the 3-stage SRVs exhibited signs of unacceptable leakage which resulted in two separate outages that involved changing out four SRVs during the first outage and the remaining seven SRVs during the subsequent outage in May 2012. The 3-stage SRVs were replaced with 2-stage SAVs containing pilot discs that had undergone the platinum surface treatment.

LEA 1-201 0*001, identified multiple SRV setpoint drift for 5 of the 11 SRVs. Corrective actions included refurbishment of the pilot valves and included the replacement of the pilot discs with discs made from Stellite 21 material. Additionally, the insulation surrounding each SRV was upgraded to improve resistance to corrosion-induced bonding. These were the same actions that were taken following similar failures reported in LEA 2-2009-001, since improved results had been seen to some degree in the industry for at least one operating cycle when these actions were implemented.

NRC FORM 36GA 102*2014)