05000456/LER-2012-001, Regarding Two Main Steam Safety Valves Failed Pre-outage Setpoint Testing Due to Abnormal Spring Geometry

From kanterella
(Redirected from 05000456/LER-2012-001)
Jump to navigation Jump to search
Regarding Two Main Steam Safety Valves Failed Pre-outage Setpoint Testing Due to Abnormal Spring Geometry
ML12163A528
Person / Time
Site: Braidwood 
(NPF-072)
Issue date: 06/11/2012
From: Enright D
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BW120055 LER 12-001-00
Download: ML12163A528 (5)
v • d • e


LER-2012-001, Regarding Two Main Steam Safety Valves Failed Pre-outage Setpoint Testing Due to Abnormal Spring Geometry
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)
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)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)(viii)(B)
10 CFR 50.73(a)(2)(ix)(A)
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
4562012001R00 - NRC Website
v • d • e

text

Exelon Generation Company, LLC Braidwood Station 35100 South Route 53, Suite 84 Braceville, IL 60407-9619 June 11,2012 BW120055 www.exeloncorp.com 10 CFR 50,73 U, S, Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No, NPF-72 NRC Docket No, STN 50-456

Subject:

Licensee Event Report 2012-001 Two Main Steam Safety Valves Failed Pre-outage Setpoint Testing Due to Abnormal Spring Geometry The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50,73, "Licensee event report system,"

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Chris VanDenburgh, Regulatory Assurance Manager, at (815) 417-2800, Respectfully, Daniel J, Enright Site Vice President Braidwood Station

Enclosure:

LER 2012-001-00 cc: NRR Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)

Illinois Emergency Management Agency - Braidwood Rep

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150*0104 EXPIRES: 10/31/2013 (10*2010)

Estimated burden to this mandatory collection request: 80 are incorporated into the

.process and back tOe~~;tJf-=

comments regarding burden to the U.S. Nuclear Regulatory or by e*mail to LICENSEE EVENT REPORT (LER)

NA" to the'Desk Officer.

of Information land la.

,,~Xn lNANN H

Office and IB~?g~t DC 20503. If means'is~d to (See reverse for required number of

'::;l~

  • does a currently control tha klQr' ~.

digits/characters for each block) inot condu~~~r';f.~~sor.

and a person is not required to respond to, the

13. PAGE Braidwood Station, Unit 1 05000456 1 of 4
4. TITLE Two Main Steam Safety Valves Failed Pre-outage Setpoint Testing Due to Abnormal Spring Geometry
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED YEAR ISEQUENTIALI REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.

MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 04 12 2012 2012 - 001 -

00 06 11 2012 N/A N/A

9. 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)

[8J 50.73(a)(2)(vii) 1 o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viil)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1)(I)(A) o 50.73(a)(2)(iil) o 50.73(a)(2)(ix)(A)

10. POWER LEVEL o 20.2203(a)(2)(li) o 50.36(c)(1)(ii)(A) o 50.73(a)(2)(lv)(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) 094 o 20.2203(a)(2)(lv) o 50.46(a)(3)(il) 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)

[8J 50.73(a)(2)(I)(B) o 50.73(a)(2)(v)(D)

Specify in Abstract below or in Contributing causes were determined to be spindle wear and steam leakage across the disc seat area. The 1MS015D valve spindle showed significant vibration fretting damage from main steam system flow resonance.

Additionally, the Inconel disc showed eight separate leak paths across the surface, which lowers the lift setpoint because it increases the huddling chamber pressure which provides lifting force during valve actuation. The 1MS014D valve Inconel disc showed two locations of significant steam leakage across the seat area in two areas 180 degrees apart which can lower the lift setpoint.

D.

Safety Consequences

There were no actual safety consequences impacting plant or public safety as a result of this event. This event captures a setpoint discrepancy, and not an actual demand for the 1MS015D and 1MS014D valves to lift.

From the surveillance testing:

1MS015D - Lifted low at 1143.79 psig (-5.08 %)

1MS014D - Lifted low at 1179.64 psig (-3.31 %)

The MSSVs are credited in the UFSAR Chapter 15 analyses for overpressure protection and small break loss of coolant accidents (SBLOCA). For the overpressure protection cases, the MSSVs are modeled with a +4 percent tolerance, and a +5 percent tolerance is used for SBLOCA - the analyses assume the MSSVs lift at pressures higher than the nominal setpoint. Lower MSSV lift pressure will be beneficial (that is gain margin to the acceptance criteria) for overpressure and the SBLOCA analyses. Since both the 1MS015D and 1MS014D lifted below their nominal setpoints, there is no impact on the overpressure and SBLOCA events.

The MSSVs are also modeled in the UFSAR Chapter 15 analyses as a release path for radiological releases.

From this event, the most limiting as-found setpoint was the 1MS015D valve which lifted at 1143.79 psig. Adding a bounding instrument uncertainty of -6.2 psig, the resultant lift pressure is 1137.59 psig (1143.79 - 6.2 =1137.59 psig). For the steam generator tube rupture (SGTR) event, the steam generator power operated relief valves (PORVs) as well as the MSSVs are modeled in this analysis. The majority of the release from this event is through the steam generator PORV, which is assumed to be stuck open for 20 minutes. A depressurization occurs as the steam generator PORV becomes stuck open. The steam generator PORV is modeled to open at a pressure of 1099.7 psia. Since this is below the 1MS015D as-found lift pressure, the MSSV will not lift and there is no impact on this accident.

Dose calculations for releases other than SGTRs assume MSSVs as a release path. The current calculation assumes all 20 MSSVs remain open at a pressure of 1022.2 psig. The MSSV test results for 1MS015D show a lift pressure of 1137.59 (1143.79 - 6.2) psig. Assuming 10 % blowdown, this one valve would stay open at a pressure of 1137.59 X 0.9 = 1023.83 psig. This value is higher than the value of 1022.2 psig assumed in the analysis, therefore, there is no impact on this accident.

Based on the discussion above, the safety function of the MSSVs was not lost. Therefore, there was no safety system functional failure due to this event.

g.

Corrective Actions

Corrective actions included refurbishing the 1MS015D and 1MS014D valves (this action is complete).

F.

Previous Occurrences

2. DOCKET 05000456 YEAR 2012
6. LER NUMBER 001 REV NO_

00 4

3. PAGE OF 4

No previous, similar Licensee Event Reports were identified at the Braidwood Station:

~.

Component Failure Data

Manufacturer DresserNomenclature Main Steam Safety Relief Valves Model 3707R Mfg. Part Number N/A

Retrieved from "https://kanterella.com/w/index.php?title=05000456/LER-2012-001,_Regarding_Two_Main_Steam_Safety_Valves_Failed_Pre-outage_Setpoint_Testing_Due_to_Abnormal_Spring_Geometry&oldid=5257557"