05000373/LER-2018-004, Technical Specification Required Shutdown Due to Reactor Pressure Boundary Leakage
| ML18141A637 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/21/2018 |
| From: | Vinyard H Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA18-037 LER 2018-004-00 | |
| Download: ML18141A637 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 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)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii) |
| 3732018004R00 - NRC Website | |
text
Exelon Generation RA18-037 May21,2018 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 1 Renewed Facility Operating Licenses No. NPF-11 NRC Docket No. 50-373 LaSalle County Station 2601 North 21 1 Road Marseilles, IL 61341 815-415-2000 Telephone www.exeloncorp com 10 CFR 50.73
Subject:
Licensee Event Report 2018-004-00, Technical Specification Required Shutdown due to Reactor Pressure Boundary Leakage In accordance with 1 O CFR 50.73(a)(2)(i)(A) and 1 O CFR 50.73(a)(2)(ii)(A), Exelon Generation Company, LLC (EGC) is submitting Licensee Event Report (LER)
Number 2018-004-00 for LaSalle County Station, Unit 1.
There are no regulatory commitments in this letter. Should you have any questions concerning this report, please contact Mr. Guy V. Ford, Jr., Regulatory Assurance Manager, at (815) 415-2800.
v;Jµ Harold T. Vinyard Plant Manager LaSalle County Station
Enclosure:
Licensee Event Report cc:
Regional Administrator-NRC Region Ill NRC Senior Resident Inspector-LaSalle County Station
NRC FORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
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3.Page LaSalle County Station, Unit 1 05000373 1
OF 3
- 4. Title Technical Specification Required Shutdown due to Reactor Pressure Boundary Leakage
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved I
Sequential I Rev Facility Name Docket Number Month Day Year Year Number No.
Month Day Year NA NA Facility Name Docket Number 03 22 18 2018 -
004 00 05 21 18 NA NA
- 9. Operating Mode
- 11. This Report Is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that apply)
D 20.2201(b)
D 20.2203(a)(3)(i)
[gj 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B) 1 D
D D
D 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(iii) 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) 24 D 20.2203(a)(2)(v)
[gj 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)
D 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)
D Other (Specify in Abstract below or in NRC Fonn 366A)
- 12. Licensee Contact for this LER Licensee Contact relephone Number (Include Area Code)
Jeff Stovall, Operations Director (815) 415-3100 Cause System Component Manufacturer Reportable to ICES
Cause
System Component Manufacturer Reportable to ICES x
AD v
A585 Yes NA NA NA NA NA
- 14. Supplemental Report Expected Month Day Year 0 Yes (If yes, complete 15. Expected Submission Date) [gj
- 15. Expected Submission Date No NA NA NA Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)
On March 22, 2018, a through-wall pipe leak was identified on a three-fourths inch vent line coming off the bonnet on a reactor recirculation pump discharge valve. This condition was reactor coolant system (RCS) pressure boundary leakage, which required entry into technical specifications (TS) actions for required shutdown. The leakage was significantly less than the 10 gallons per minute emergency action threshold. At the time of discovery, Unit 1 was in Mode 1. A controlled shutdown of Unit 1 was performed, having all necessary shutdown equipment available; and, there was no impact to Unit 2. This condition was reported on March 22, 2018 (ENS 53276) in accordance with 10 CFR 50.72(b)(2)(i) for plant shutdown required by Technical Specifications and 10 CFR 50.72(b)(3)(ii)A for the pressure boundary leakage as a principal safety barrier being in a degraded condition.
The cause for the steam leak on the vent line was determined to be fatigue cracking that initiated on the external surface of the pipe at the toe of the socket fillet weld. Crack propagation occurred under low cycle fatigue conditions. In addition, the fracture surface contained deformation features which were indicative of a potential high nominal stress. Actions were taken to repair the vent line after plant shutdown and perform failure analysis on the degraded component. This condition is reportable in accordance with 10 CFR 50.73(a)(2)(i)A as a condition that required a plant shutdown in accordance with TS and 10 CFR 50.73(a)(2)(ii)A as a principal safety barrier being in a degraded condition.
NRC FORM 366 (04-2018)
PLANT AND SYSTEM IDENTIFICATION
SEQUENTIAL NUMBER 004 REV NO.
00 LaSalle County Station Unit 1 is a General Electric Boiling Water Reactor with 3546 Megawatts Thermal Rated Core Power.
The affected component was a three-fourths inch vent line (1 RR32AB-3/4") coming off the bonnet on the 1 B reactor recirculation pump discharge valve (1833-F0678).
CONDITION PRIOR TO EVENT
Unit(s): 1 Date:
Reactor Mode(s): 1 Mode(s) Name:
DESCRIPTION
March 22, 2018 Power Operation Time:
Power Level:
0300 CDT 24 percent On March 22, 2018, at 0300 CDT, a through-wall pipe leak was identified on a three-fourths inch vent line coming off the bonnet on the 1 B reactor recirculation pump discharge valve (1833-F0678). Personnel observed a plume of steam was coming from a partial circumferential crack at the toe of the weld connecting the small-bore vent pipe 1 RR32AB-3/4" and the valve bonnet.
This condition was determined as a reactor coolant system (RCS) pressure boundary leakage, which required entry into technical specifications (TS) actions for required shutdown. Specifically, TS 3.4.5, Reactor Coolant System Operational Leakage, Required Actions C.1 and C.2 were entered, which require the unit to be in Mode 3, Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and Mode 4, Cold Shutdown in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively.
The leakage was significantly Jess than the 10 gallons per minute emergency action threshold. At the time of discovery, Unit 1 was in Mode 1. A controlled shutdown of Unit 1 was performed, having all necessary shutdown equipment available. Unit 1 was placed into Mode 4 on March 22, 2018, at 1652 CDT. There was no impact to Unit 2. This condition was reported to the NRC via Event Notification System (ENS) 53276 on March 22, 2018 as a degraded condition that involved a required plant shutdown.
CAUSE
The failure occurred after approximately seven days of service following the vent line installation during the Unit 1 refueling outage (L 1R17). A component failure analysis was completed which determined that the failure was attributed to fatigue cracking that initiated on the external surface of the pipe at the toe of the socket fillet weld. Based on the relatively large spacing of the fracture surface fatigues striations, crack propagation occurred under low cycle fatigue conditions. In addition, the fracture surface contained deformation features which were indicative of a potential high nominal stress.
For the failed "B" vent line assembly, the likely cause for the high nominal stress was contact with the clamshell support clamp.
Based on the clamshell clamp and pipe wear marks, the support appeared to have applied a downward force to the failed line during service. Lab inspections did not detect fatigue cracking at the socket weld location in the similar intact "A" vent line, which suggests that failure was not imminent.
The station initiated a root cause investigation to determine additional causal factors for this failure.
REPORTABILITY AND SAFETY ANALYSIS
This condition is reportable in accordance with 10 CFR 50.73(a)(2)(i)A as a condition that required a plant shutdown in accordance with TS and 10 CFR 50.73(a)(2)(ii)A as a principal safety barrier being in a degraded condition. This condition was reported to the NRC via ENS 53276 on March 22, 2018.
There were no safety consequences resulting from the event. Makeup capability was adequate to compensate for the leak. All emergency core cooling systems (ECCS) were operable and capable of fulfilling their intended safety functions during the period of leakage. The leak was within the capability of the leak detection system and was within the capability of the leak collection system.
YEAR 2018 SEQUENTIAL NUMBER 004 The event did not constitute a safety system function failure (SSFF) as defined in accordance with NEI 99-02, "Regulatory Assessment Performance Indicator Guideline."
CORRECTIVE ACTIONS
Immediate corrective actions taken in response to the condition were:
Performed required plant shutdown Removed the degraded vent line and valves from 1833-F0678 Removed the vent lines and valves from 1833-F067 A Performed failure analysis of the failed component Initiated a causal investigation Additional corrective actions may be determined from the root cause investigation.
PREVIOUS OCCURRENCES
LER 374-2015-003-00 REV NO.
00 On August 7, 2015, LaSalle Unit 2 was in Mode 3 for a planned maintenance outage. During the initial drywell entry, a steam leak was observed on the reactor recirculation (RR) system line 2RR94AB-3/4", which is upstream of valve 2833-FOSOB (RR Pump Discharge Valve 2833-F0678 Inspection Port - Reactor Side Upstream Stop Valve). The leak was determined to be pressure boundary leakage. Technical Specification 3.4.5, "RCS Operational Leakage," Required Actions C.1 and C.2 were entered, which required the unit to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, respectively. The cause of the steam leak was determined to be poor weld quality and vibration induced fatigue. The weld was repaired during the maintenance outage.
COMPONENT FAILIRE DATA Manufacturer: Atwood & Morrill (A585)
Device: Reactor Recirculation Pump Discharge Valve 1833-F0678 Component ID: Model Series 9085 Wedge Gate Valve Page _3_ of _3_