05000388/LER-2005-001, Regarding Degradation of Primary Coolant Pressure Boundary Due to Recirculation Pump Discharge Valve Bonnet Vent Connection Weld Flaw
| ML051440352 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  (NPF-022) |
| Issue date: | 05/12/2005 |
| From: | Saccone R Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-5893 LER 05-001-00 | |
| Download: ML051440352 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3882005001R00 - NRC Website | |
text
Robert A. Saccone PPL Susquehanna, LLC I I
- Vice President - Nuclear Operations 769 Salem Boulevard
% & a a '
Berwick, PA 18603 N45s" Tel. 570.542.3698 Fax 570.542.1504 rasaccone~pplweb.com MAY 1 R 2ifl5 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OPI-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-388/2005-001-00 LICENSE NO. NPF-22 PLA-5893 Docket No. 50-388 Attached is Licensee Event Report (LER) 50-388/2005-001-00. This event was determined to be reportable per 10 CFR 50.73(a)(2)(ii)(A).
On March 20, 2005, during the performance of the Unit 2 ASME Class 1 Boundary Leakage.
Test, a socket weld at the connection to the bonnet vent piping for the "B" Reactor Recirculation Pump Discharge Valve HV243F031B was found leaking approximately 10 drops per minute. Subsequent investigation revealed a crack in the toe of the weld at the valve's bonnet vent socket weld. The connection was an abandoned piece of 3/4-inch piping that contained a 16-inch long pipe stub that had been cut and capped in 1997 due to a previous vibration-induced weld failure. This condition constituted a degradation of the reactor coolant pressure boundary, and on March 21, 2005, an ENS Notification (#41506) was made to the NRC.
This event resulted in no actual adverse consequences to the health and safety of the public.
No commitments are associated with this LER.
Vice rsident - Nuclear Operations Attachment Document Control Desk PLA-5893 cc:
Mr. S. Collins Regional Administrator U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. A. Blarney Sr. Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 35 Berwick, PA 18603-0035 Mr. R. Osborne Allegheny Electric Cooperative P. 0. Box 1266 Harrisburg, PA 17108-1266 Mr. R. R. Janati Bureau of Radiation Protection Rachel Carson State Office Building P. 0. Box. 8469 Harrisburg, PA 17105-8469
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB: NO,.3150-0104 EXPIRES:O6=r2O07 (6-20134)Estimated
, the NRC may not conduct or sponsor, and a person is not required to respond to, the of digits/characters for each block)
Information collection.
- 1. FACILlTY NAME Susquehanna Steam Electric Station Unit 2 DOCKET NUMBER
- 3. PAGE 05000388
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1 OF3
- 4. TITLE Degradation of Primary Coolant Pressure Boundary due to Recirculation Pump Discharge Valve Bonnet Vent Connection Weld Flaw
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILmES INVOLVED MONTH DAY YEAR SEQUENTIALI REV MONTH DAY YEAR FACIUTY NAME DOCKET NUMBER NUMBER NO.
FACILITY NAME DOCKET NUMBER 3
20 2005 01 00 5
12 2005
- 9. OPERATING MODE 11.THIS REPORTIS SUBM=ED PURSUANI TTOTHE REQUIREMENTS OF 10 CFR §: (Check all that apply) 4 0 20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii)
- 10. POWER LEVEL 0 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A)
Eo 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 0%
0 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A) o 20.2203(a)(2)(iqi 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) 0 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71 (a)(4) o 20.2203(a)(2)(v) 0 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) 0 73.71 (a)(5) 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER 0 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)
Specify in Abstract below or In (If more space Is required use addtimalccpies ofNRCForm 3)
PLANT CONDITIONS AT TIME OF EVENT Unit 1, Mode 1, 100%
Unit 2, Mode 4, 0%
EVENT DESCRIPTION
On March 20, 2005, during the performance of the Unit 2 ASME Class 1 Boundary Leakage Test, a socket weld at the connection to the bonnet vent piping for the "B" Reactor Recirculation Pump Discharge Valve HV243F031 B (EIIS: AD) was found leaking approximately 10 drops per minute. Subsequent investigation revealed a crack in the toe of the weld at the valve's bonnet vent socket weld. The connection was an abandoned piece of 3/4-inch piping that contained a 16-inch long pipe stub that had been cut and capped in 1997 due to a previous vibration-induced weld failure.
This condition constituted pressure boundary leakage as defined in Technical Requirement 3.4.2 and TRO Action 3.4.2.C was entered. On March 21, 2005, an ENS Notification (#41506) was made to the NRC. Based on guidance provided in NUREG-1 022, Rev. 2, this material defect in the primary coolant boundary is unacceptable under ASME Code requirements and is reportable under 10 CFR 50.73(a)(2)(ii)(A).
CAUSE OF THE EVENT
The cause of the weld failure was due to inadequate implementation of GE SIL 512 recommendations. GE SIL 512 identified preventative measures for socket welds associated with cantilevered branch lines attached to recirculation gate valves. The GE SIL recommended plugging the branch lines or minimizing the relative motion of the branch line by strapping the line to the recirculation pipe. In 1997, PPL cut the 3/4-inch line and capped it leaving a 16-inch long pipe stub. However, the GE SIL recommended removing the line at the point of highest fatigue and plugging it.
ANALYSIS / SAFETY SIGNIFICANCE Actual Consequences There were no safety consequences or compromises to public health and safety as a result of the failed weld.
With the reactor shutdown in Mode 4, the available systems were fully capable of compensating for the leakage.
Based on the location of the flaw, the leak path was also capable of being isolated.
Potential Consequences In Mode 4, a complete weld failure would have resulted in an isolation of the leak path and a controlled vessel depressurization by operators. Auxiliary systems, such as control rod drive and low pressure emergency core cooling systems, were also available to maintain vessel inventory, if required.
NRC FORM 366 (6-2004)
(If mno space Is emuired, use ad cnalcqoies ofNRG F&m 366A)
During Mode 1 operation, the weld flaw would have resulted in increased drywell unidentified leakage and an eventual failure of the 1 6-inch long pipe stub. This loss of reactor vessel inventory is bounded by the PPL analysis as described in Chapter 15 of the FSAR. Small pipe break LOCA events, including breaks on this particular line, have been evaluated as part of PPL's PRA model. These events contribute less than 2% of the risk of core damage frequency (CDF) and are evaluated as an insignificant contributor (<1.OE-05%) to the large early release frequency. This particular line makes up approximately 10% of the total small liquid LOCA event frequency; therefore, the CDF for this particular line break would contribute less than 0.2% of the overall CDF.
CORRECTIVE ACTIONS
Completed Actions The cracked weld was repaired by removing the pipe stub and welding a plug directly into the bonnet connection.
A review of other large bore ASME Class 1 gate valves (not associated with the GE SIL) inside Unit 1 and Unit 2 containment was performed to determine if similar small pipe connections existed. No similar connections were found.
Planned Actions A new review of Unit 1 and Unit 2 recirculation lines identified by GE SIL 512 will be performed. All recirculation gate valve vent, drain, and stem leak-off lines, previously abandoned will be plugged at the welded joint. Other vibration-prone small bore piping will also be evaluated for susceptibility to weld cracking and corrective actions taken as necessary to prevent future weld failures.
ADDITIONAL INFORMATION
Past Similar Events:
Docket No. 50-388, LER 93-009 and LER 97-006 NRL FURM 366 (6-2004)