05000321/LER-2016-004, Safety Relief Valves as Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria
| ML16147A344 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/26/2016 |
| From: | Pierce C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-16-0772 LER 16-004-00 | |
| Download: ML16147A344 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3212016004R00 - NRC Website | |
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Charles R. Pierce Regulatory Affairs Director May 26,2016 Docket Nos.: 50-321 Southern Nuclear Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 Fax 205.992.7601 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report 2016-004-00 SOUTHERN<<\\
NUCLEAR A SOUTHERN COMPANY NL-16-0772 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. If you have any questions, please contact Greg Johnson at (912) 537-5874.
Rtecflft; C. R. Pierce Regulatory Affairs Director CRP/cp/lac Enclosures: LER 2016-004-00
U.S. Nuclear Regulatory Commission NL-16-0772 Page2 cc:
Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Best, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President-Hatch Mr. M. D. Meier, 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 Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager-Hatch Mr. D. H. Hardage, Senior Resident Inspector-Hatch
Edwin I. Hatch Nuclear Plant Unit 1 LER 2016-004-001 Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150*0104 EXPIRES: 10/31/2018 (11*2015),.......
, the NRC mav not conduct or sponsor and a oerson is not reauired to resoond to the inlormation
- 1. FACILITY NAME
~ - DOCKET NUMBER j*
PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000 321 1 OF4
- 4. TITLE Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV MONTH FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
DAY YEAR FACILITY NAME DOCKET NUMBER 3
30 2016 2016
. 004
- - 00 5
26 2016
- 9. OPERATING MODE
- 11. THIS REPORTIS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.22o1(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 1 D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 100 D 20.2203(a)(2)(iv)
D 50.46(a)(3)(il)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi) 181S0.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 00THER Specify in Abstract below or in There are eleven SRVs located on the four main steam lines within the drywell in between the reactor pressure vessel (RPV) (EllS Code RPV) and the inboard main steam isolation valves (MSIVs) (EllS Code ISV). 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 SRVs 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).
The two SRVs which failed to meet their Tech Spec required actuation pressure setpoint lifted early (3.2% low and 3.8% low).
None of the eleven SRVs tested this cycle had as-found test results out of range high. Therefore, since the two identified SRVs lifted earlier than expected, the ASME Code Limit of 1375 psig peak vessel pressure would be maintained under normal and accident conditions. The opening of one or more SRVs at lower pressures would result in a less severe transient with reduced peak vessel pressure. Also, the slightly lower actuating pressure does not pose a significant LOCA initiator threat because the reactor steam dome would not experience > 11 00 psig during normal operation.
Based on the observed setpoint drift slightly low, 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 vendor specifications will be revised to tighten as-left tolerances of abutment and pre-load gap, increase the minimum set for abutment pressure at the high end of specification, and tighten diametrical and face run-out tolerances for bellows assembly on pre-load spacer mounting end.
ADDITIONAL INFORMATION
Other Systems Affected: None
Failed Components Information
Master Parts List Number: 1 821-F013D, E Manufacturer: Target Rock Model Number: 0867F Type: Relief Valve Manufacturer Code: T020 EllS System Code: SB Reportable to EPIX: Yes Root Cause Code: 8 EllS Component Code: RV 05000-321 Commitment Information: This report does not create any licensing commitments.
PREVIOUS SIMILAR EVENTS
YEAR I 2016 SEQUENTIAL NUMBER
- - 004 I
REV NO.
- - 00 LER 2-2015-004 identified multiple SRV setpoint drift for 2 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs which typically do not exhibit set point drift. The setpoint drift was out of spec high while the event discussed in LER 1-2016-004 have failed to meet acceptance criteria by drifting out of spec low.
LER 1-2014-003 identified multiple SRV setpoint drift for 5 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs which typically do not exhibit set point drift. The setpoint drift was out of spec high while the event discussed in LER 1-2016-004 have failed to meet acceptance criteria by drifting out of spec low.
LER 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.
Charles R. Pierce Regulatory Affairs Director May 26,2016 Docket Nos.: 50-321 Southern Nuclear Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 Fax 205.992.7601 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report 2016-004-00 SOUTHERN<<\\
NUCLEAR A SOUTHERN COMPANY NL-16-0772 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. If you have any questions, please contact Greg Johnson at (912) 537-5874.
Rtecflft; C. R. Pierce Regulatory Affairs Director CRP/cp/lac Enclosures: LER 2016-004-00
U.S. Nuclear Regulatory Commission NL-16-0772 Page2 cc:
Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Best, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President-Hatch Mr. M. D. Meier, 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 Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager-Hatch Mr. D. H. Hardage, Senior Resident Inspector-Hatch
Edwin I. Hatch Nuclear Plant Unit 1 LER 2016-004-001 Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150*0104 EXPIRES: 10/31/2018 (11*2015),.......
, the NRC mav not conduct or sponsor and a oerson is not reauired to resoond to the inlormation
- 1. FACILITY NAME
~ - DOCKET NUMBER j*
PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000 321 1 OF4
- 4. TITLE Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV MONTH FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
DAY YEAR FACILITY NAME DOCKET NUMBER 3
30 2016 2016
. 004
- - 00 5
26 2016
- 9. OPERATING MODE
- 11. THIS REPORTIS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.22o1(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 1 D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 100 D 20.2203(a)(2)(iv)
D 50.46(a)(3)(il)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi) 181S0.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 00THER Specify in Abstract below or in
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 30 2016, with Unit 1 at 100 percent rated thermal power (RTP), "as-found" testing of the 3-stage main steam safety relief valves (SRVs) (EllS Code RV) showed that two of the eleven main steam SRVs that were tested had experienced a drift in pressure lift setpoint during the previous operating cycle such that the allowable technical specification (TS) surveillance requirement (SR) 3.4.3.1 limit of 1150 +1-34.5 ( +1-3%) psig had been exceeded. Below is a table illustrating the Unit 1 SRVs that failed as found testing results after being removed from service during the Spring 2016 refueling outage.
MPL 1821-F013D 1B21-F013E
CAUSE OF EVENT
Lift Pressure 1113 psig 1106 psig Percent Drift
- - 3.2%
- - 3.8%
The SRV pilots were disassembled and inspected while investigating the reason for the drift. SNC has determined that the linear variable differential transformer (LVDT) used to measure displacement confirmed the abutment gap closed pre-maturely. The pre-mature abutment gap closure is most likely due to loose manufacturing tolerances leading to SRV setpoint drift.
REPORT ABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable in accordance with 10 CFR 50. 73(a)(2)(i)(B) because a condition occurred that is prohibited by TS 3.4.3. Specifically, an example of multiple test failures is given in NUREG-1 022, 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, the determination was made that this "as found" condition is reportable under the reporting requirements of 10 CFR 50.73(a)(2)(i)(B).
There are eleven SRVs located on the four main steam lines within the drywell in between the reactor pressure vessel (RPV) (EllS Code RPV) and the inboard main steam isolation valves (MSIVs) (EllS Code ISV). 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 SRVs 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).
The two SRVs which failed to meet their Tech Spec required actuation pressure setpoint lifted early (3.2% low and 3.8% low).
None of the eleven SRVs tested this cycle had as-found test results out of range high. Therefore, since the two identified SRVs lifted earlier than expected, the ASME Code Limit of 1375 psig peak vessel pressure would be maintained under normal and accident conditions. The opening of one or more SRVs at lower pressures would result in a less severe transient with reduced peak vessel pressure. Also, the slightly lower actuating pressure does not pose a significant LOCA initiator threat because the reactor steam dome would not experience > 11 00 psig during normal operation.
Based on the observed setpoint drift slightly low, 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 vendor specifications will be revised to tighten as-left tolerances of abutment and pre-load gap, increase the minimum set for abutment pressure at the high end of specification, and tighten diametrical and face run-out tolerances for bellows assembly on pre-load spacer mounting end.
ADDITIONAL INFORMATION
Other Systems Affected: None
Failed Components Information
Master Parts List Number: 1 821-F013D, E Manufacturer: Target Rock Model Number: 0867F Type: Relief Valve Manufacturer Code: T020 EllS System Code: SB Reportable to EPIX: Yes Root Cause Code: 8 EllS Component Code: RV 05000-321 Commitment Information: This report does not create any licensing commitments.
PREVIOUS SIMILAR EVENTS
YEAR I 2016 SEQUENTIAL NUMBER
- - 004 I
REV NO.
- - 00 LER 2-2015-004 identified multiple SRV setpoint drift for 2 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs which typically do not exhibit set point drift. The setpoint drift was out of spec high while the event discussed in LER 1-2016-004 have failed to meet acceptance criteria by drifting out of spec low.
LER 1-2014-003 identified multiple SRV setpoint drift for 5 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs which typically do not exhibit set point drift. The setpoint drift was out of spec high while the event discussed in LER 1-2016-004 have failed to meet acceptance criteria by drifting out of spec low.
LER 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.