05000364/LER-1917-002, Regarding Main Steam Safety Valve Lift Pressure Outside of Technical Specifications Limits
| ML17353A931 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/19/2017 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-17-2107 LER 17-002-00 | |
| Download: ML17353A931 (5) | |
| 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)(iii) 10 CFR 50.73(a)(2)(ix) 10 CFR 50.73(a)(2) 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) |
| 3641917002R00 - NRC Website | |
text
~ Southern Nuclear DEC 1 9 2017 Docket No.:
50-364 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D. C. 20555-0001 Dennis R. Malfson Vu:e President-Farley Joseph M. Farley Nuclear Plant - Unit 2 Ucensee Event Report 2017-002-00 Main Steam Safety Valve Lift Pressure Outside of Technical Specifications Limits Ladies and Gentlemen:
Jasqil M. F.anlcy Nuckar 7388 Kulh SU2e HU) 95 CC!inm!>Q. Ab!=na.J6..119 33.UI-1.4!iiJ llel 3JUIIU575 fu NL-17-2107 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 2017-002-00 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 2017-002-QO Main Steam Safety Valve Lift Pressure Outside of Technical Specifications Limits
NRC FORII 366 U.S. NUCLEAR REGULATORY COIIIIISSION APPROVED BYOIIB: NO. 31~04 EXPIRES: 0313112020 (04-Zn1) fsfima!ed tu!len ;er fl!!illliiSI! tl ~
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- 1. FACIUTY NAME
- 2. DOCKET NUMBER
- 3. PAGE Joseph M. Farley Nuclear Plant, Unit 2 05000 364 I OF 3
- 4. TITLE Main Steam Safety Valve Lift Pressure Outside ofTechnical Specifications Limits
- 5. EVENT DAlE
- 6. LER NUMBER
- 7. REPORT DAlE
- 8. OTHER FACILmES INVOLVED I
SEQUENTIAL I REV FACUJTY NAME llOCI<ET NUiolllER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NO.
05000 FACUJTYNAIIE llOCI<ET NIJMI!ER II 01 2017 2017 -
002 -
00 12 19 2017 05000
- 9. OPERAnNG MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 2o.22o1(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 6 D 2o.22o1(d)
D 20.2203(a)(3)(ii) 0 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)
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1) 000 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) 0 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C)
OoTHER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER UCENSEE CONTACT TELEPHONE NUMBER (lndude Atea COde)
Mandy Ludlam, Licensing Engineer (334) 814-4930 CAUSE SYSTEM COMPONENT MANU*
REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX E
SB RV 0245 y
t
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO SUBMISSION DATE ABSTRACT (Umit to 1400 spaces, i.e., approximately 15 single-spaced typewritten fines)
On November I, 2017, while in Mode 6 and at 0% power level, one ofthe C Loop Main Steam Safety Valves (MSSV) as-found lift pressure did not meet the acceptance criteria of+/- 3% of the setpoint (I 129 psig) as required by Technical Specifications (TS)
Surveillance Requirement (SR) 3. 7. I. I. The MSSV lifted at I I 7 I psig which is 9 psig outside of its acceptance range of I 096 to I I 62 psig and 3.72% above its setpoint. The apparent cause of exceeding the MSSV upper acceptance limit is degradation of the valve spring and/or valve spindle compression screw. The as-found settings remained within analytical bounds; therefore, operation of the facility in this condition had no impact on the health and safety of the public.
TS Limiting Condition for Operation (LCO) 3.7.1, MSSVs, requires five MSSVs per steam generator to be operable in Modes I, 2, and
- 3. Since the failure affected the lift pressure over a period oftime, it is assumed that the C Loop MSSV was inoperable for a time greater than allowed by TS. Therefore, this occurrence is considered reportable per 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited byTS.
The C Loop MSSV was replaced on November 5, 2017, while in Mode 5.
NRC FORM 386 (04-2017)
NRC FORII366A (0'-2017)
U.S. NUCLEAR REGULATORY COIIMJSSIOH APPROVED BY 0118: NO. 3150-0104 EXPIRES: 0313112020 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022. R.3 for instruction and guidance for completing this fonn
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- 1. FAQUTY NAME
- 2. DOCKET NUMBER 3.LER NUMBER Joseph M. Farley Nuclear Plant, Unit 2 05000-1 364 I
YEAR SEQUENTIAL REV 1
2017 1-1 N= 1-1 : I NARRAT1VE
EVENT DESCRIPTION
On November 1, 2017, while in Mode 6 and at 0% power level for refueling outage 2R25, with the Reactor Coolant System (RCS) at atmospheric pressure and 84 degrees Fahrenheit, one of the C Loop Main Steam Safety Valves (MSSV) as-found lift pressure did not meet the acceptance criteria of+/- 3% of setpoint (1129 psig) as required by Technical Specifications (TS) Surveillance Requirement (SR) 3.7.1.1 when tested by an off-site testing facility per their testing guidelines and in accordance with plant procedures. The MSSV lifted high at 1171 psig which is 9 psig outside of its acceptance range of 1096 to 1162 psig and 3.72% above its setpoint. The+/- 3% as-found lift pressure requirement is an ASME Section Ill, 1971 edition, and Farley Technical Specification (TS) requirement to ensure that the MSSV provides adequate protection by preventing the steam pressure from exceeding 11 0 percent of the main steam system design pressure.
EVENT CAUSE ANALYSIS
Evaluation of failures of a B Loop MSSV in 201 0 and 2013 concluded there was inadequate preventive maintenance inspections on the MSSVs. As a result, a 10.5-year inspection and refurbishment preventative maintenance task was created with a comprehensive plan to inspect and replace the subcomponent in all MSSVs. Each MSSV was scheduled for inspection and replacement of valve disk/material and spindle compression screw assembly with dampened vibration.
The C Loop MSSV was the last remaining valve to be inspected and rebuilt as part of this comprehensive plan. The valve removed from the C Loop MSSV location had been tested satisfactorily prior to refueling outage 2R24 and was not part of the In-Service Testing (1ST) scope for 2R25. It was removed and tested in 2R25 as part of the comprehensive plan. The apparent cause of exceeding the MSSV upper acceptance limit is degradation of the valve spring and/or valve spindle compression screw. The as-found settings remained within analytical bounds; therefore, operation of the facility in this condition had no impact on the health and safety of the public.
REPORTABLITY AND SAFETY ASSESSEMENT:
This event is reportable in accordance with 10CFR50.73(a)(2)(i)(B). The applicable accidenUtransient analyses requires five MSSVs per Steam Generator (SG) to provide overpressure protection for design basis transients occurring at 102%
Rated Thermal Power. The MSSVs also provide a heat sink for the RCS if the main condenser is unavailable and the atmospheric dump valves cannot relieve steam line pressure. Operability of the MSSVs is defined as the ability to open within the setpoint range, relieve SG overpressure, and re-seat when pressure has been reduced, and is determined by periodic surveillance testing. On November 1, 2017, a C Loop MSSV was found outside of its required setpoint range; therefore, it failed its as-found testing criteria and was declared inoperable. The apparent cause determined that the failure was due to degradation of the valve spring and/or valve spindle compression screw. This degradation is not normal drift; therefore, the valve may have been inoperable during past operation. As it is not possible to determine when the valve would have exceeded the setpoint range, the C Loop MSSV was determined to be inoperable for greater than the TS allowed completion time. Based on the MSSV as-found lift setpoint being less than 110% of design Steam Generator pressure (1194 psig), this one MSSV failure would not have resulted in a loss of safety function. Therefore, this condition is not reportable under 10CFR50.73(a)(2)(v) as a safety system functional failure.
CORRECTIVE ACTIONS
The C Loop MSSV was replaced on November 5, 2017, while in Mode 5.
PREVIOUS SIMILAR EVENTS
One of the 8 Loop MSSVs lifted low outside of the+/- 3% lift pressure requirement in 2010 and lifted high outside of+/-
3% lift pressure requirement in 2013. An analysis of both failures identified inadequate preventive maintenance inspections on the MSSVs. Refurbishment of the C Loop MSSV was initiated as a corrective action from the analysis of the previous events on the 8 Loop MSSV.
OTHER SYSTEMS AFFECTED:
No other systems were affected by this event. Page 3
of 3