05000364/LER-1917-003, Regarding Pressurizer Safety Valve Lift Pressure Outside of Technical Specifications Limits
| ML17354A395 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/20/2017 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-17-2109 LER 17-003-00 | |
| Download: ML17354A395 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 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 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3641917003R00 - NRC Website | |
text
~ Southern Nuclear DEC 2 0 2017 Docket No.:
50-364 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D. C. 20555-0001 Dennis R. Malison V"ICe President - Farley Joseph M. Farley Nuclear Plant - Unit 2 Licensee Event Report 2017-003-00 Pressurizer Safetv Valve Lift Pressure Outside of Technical Specifications Limits Ladies and Gentlemen:
Jos.;il M.. F~
~"ude.a-..,_
T..".SS !Nonb Stmo Huy 9S Cobmiria. A!'au!rz.l36319 3J.I.fU4.till td 33UI-1.-b15 fn NL-17-2109 In accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), Southern Nuclear Company is submitting the enclosed Licensee Event Report for Unit 2.
This letter contains no NRC commitments. If you have any questions regarding this submittal, please contact Mandy Ludlam at (334) 814-4930.
Respectfully submitted, D.R. Madison Vice President-Farley DRM/mmVcbg Enclosure: Unit 2 Licensee Event Report 2017-003-00 Cc: Regional Administrator, Region II NRR Project Manager-Farley Nuclear Plant Senior Resident Inspector-Farley Nuclear Plant RTYPE: CFA04.054
Enclosure Joseph M. Farley Nuclear Plant Unit 2 Licensee Event Report 2017-003-00 Pressurizer Safety Valve Lift Pressure Outside of Technical Specifications Limits
NRC FORII366 U.S. NUCLEAR REGULATORY COIIIIISSION APPROVED BY 0118: NO. 3'150-0104 EXPIRES: 0313112020 (04-2017}
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- 1. FACIUTY NAME
- 2. DOCKET NUMBER 3.PAGE Joseph M. Farley Nuclear Plant, Unit 2 05000 364 I OF 2
- 4. llTLE Pressurizer Safety Valve Lift Pressure Outside of Technical Specifications Limits
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OlHER FACIUTIES INVOLVED I
SEQl£NfLq I REV FAQUTYNAaE IJOCI(ET NIJUBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NO 05000 12 20 201 7 FACIIJTYNAaE IJOCI(ET NUMBER 10 31 2017 20I7 -
003 -
00 05000
- 9. OPERA TlNG MODE
- 11. lHIS REPORT IS SUBMITTED PURSUANT TO lHE REQUIREMENTS OF 10 CFR §: {Check all that apply) 0 20.2201(b>
0 20.2203(a)(3)(i) 0 50.73(a)(2)(ii}(A) 0 50.73(a)(2)(viii)(A) 6 0 2o.22o1cd) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii}(B) 0 50.73(a)(2)(viii)(B) 0 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A) 0 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x)
- 10. POWER LEVEL 0 20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(v)(A) 0 73.71(a)(4) 0 20.2203(a)(2)(iii)
D 50.36(c)(2) 0 50.73(a)(2)(v)(B) 0 73.71(a)(5) 0 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) 0 50. 73(a)(2)(v)(C) 0 73.77(a)(1) 000 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i) 0 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(vii) 0 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 00THER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER UCENSEE CONTACT TELEPHONE NUMBER (lndude Area Code)
Mandy Ludlam, Licensing Engineer (334) 814-4930 CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU*
REPORTABLE FACTURER TO EPIX FACTURER TOEPIX E
AB RV C7IO y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR DYES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO SUBMISSION DATE ABSTRACT (Umit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On October 3 I, 2017, while in Mode 6 at 0% power level, it was discovered that a Unit 2 pressurizer safety valve (PSV), which had been removed during the October 20 I 7 refueling outage (2R25) and shipped offsite for testing, failed its as-found lift pressure test. The PSV lifted below the Technical Specification (TS) 3.4. I 0 allowable lift setting value. Setpoint drift of the PSV is the most likely cause of the failure.
It is likely that the PSV was outside of the TS limits longer than allowed by the Required Action Statement (15 minutes) during the previous operating cycle in all applicable modes of operation. Therefore, this condition is being reported in accordance with I 0 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS.
The PSV was replaced during the October 20 I 7 refueling outage.
NRC FORM 366 (04-2017)
NRC FORII366A (04-21117)
U.S. NUCLEAR REGULATORY C01U1JSS10N APPROVED BY 0118: NO. 3150-0104 EXPIRES: 0313112020 UCENSEE EVENT REPORT (LER)
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- 1. FACIUTY NAIIE 2.. DOCKET NUMBER
- 3. LER NUMBER Joseph M. Farley Nuclear Plant, Unit 2 05000-1 364 I
YEAR SEQUENllAL REV 12017 1-1 N= 1-1
~~- I NARRATlVE
EVENT DESCRIPTION
During the Unit 2 October 2017 refueling outage (2R25), while in Mode 6 and at 0% power level, with the Reactor Coolant System (RCS) at atmospheric pressure and 83 degrees Fahrenheit, a pressurizer safety valve (PSV) was removed as part of the routine In-Service Testing (1ST) program and sent to an off-site testing facility. The as-found lift pressure was discovered to be 2455 psig which was outside of the Technical Specification (TS) 3.4.1 0 allowable lift pressure settings of
>/= 2460 psig and </= 2510 psig. The tested valve was in the ASME code acceptance band of+/- 3% (2411-2559 psig).
Based on the lift pressure meeting the 1ST Program (ASME code) monitored requirements, there was no 1ST scope expansion for the PSV.
EVENT ANALYSIS
During the previous cycle, indications of seat leakage from this PSV was evidenced by tailpipe temperature indication.
Based on review of trends, during the 18-month cycle the tailpipe temperatures were between the minimum and maximum values, 75 and 130 degrees respectively. Additionally, leakage past the seat was identified during testing at the off-site facility. The cause of the valve removed from PSV location lifting low at 2455 psig is setpoint drift which resulted in the seat leakage.
REPORTABILITY AND SAFETY ASSESSMENT
This failure constitutes a condition that is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications." Although seat leakage was identified while online in June of 2016 there is no firm evidence of when the failure to meet the lift setting requirements occurred prior to the time of discovery at the test facility. The setpoint could have drifted below the allowable value at any time between startup from the previous refueling outage (2R24) and the time of discovery.
Since the as-found lift setpoint was lower than the allowed value in the TS, the condition did not have an adverse impact on its over-pressurization function. The as-found lift pressure was 2455 psig, and the valve re-closed following the lift. This is within the safety analysis assumptions that are credited for PSVs, and the plant remained bounded by the accident analyses in the Final Safety Analysis Report (FSAR). Therefore, this condition had no significant effect on the health and safety of the public.
CORRECTIVE ACTIONS
The PSV was replaced during the October 2017 refueling outage. The as-left setpoints were within +/- 1% tolerance.
PREVIOUS SIMILAR EVENTS
Similar events were reported for Unit 1 in LER 2015-004-00 and LER 2016-003-00. For the 2015 LER, there had been indication of seat leakage during the previous operating cycle based on elevated tailpipe temperatures.
OTHER SYSTEMS AFFECTED:
No other systems were affected by this event. Page 2
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