05000387/LER-2014-011, Regarding Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by an Inadequate Weld
| ML15042A469 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/11/2015 |
| From: | Franke J Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-7286 LER 14-011-00 | |
| Download: ML15042A469 (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 |
| 3872014011R00 - NRC Website | |
text
FEB 11 2015 Jon A. Franke Site Vice President U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 jfranke@pplweb.com SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387/2014-011-00 UNIT 1 LICENSE NO. NPF-14 PLA-7286 Docket No. 50-387 Attached is Licensee Event Report (LER) 50-387/2014-011-00. The LER reports an event involving a degraded condition due to Reactor Coolant Pressure Boundary leakage in accordance with 10 CFR 50.73(a)(2)(ii)(A).
There were no actual consequences to the health and safety of the public as a result of this event.
This letter contains no new regulatory commitments.
Attachment: LER 387/2014-011-00 Copy:
NRC Region I Mr. J. E. Greives, NRC Sr. Resident Inspector Mr. J. A. Whited, NRC Project Manager Mr. L. J. Winker, PA DEP/BRP
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
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Reported lessons learned are incorporated into the licensing process and fed back to industry.
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LICENSEE EVENT REPORT (LER)
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T*5 F53). U.S. Nuclear Regulatory Commission, Washington, DC 20555*0001, or by (See Page 2 for required number of internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB*1 0202, (3150*01 04), Office of Management and Budget, Washington, DC digits/characters for each block) 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
r* PAGE Susquehanna Steam Electric Station, Unit 1 05000387 1 of4
- 4. TITLE Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by an Inadequate Weld
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NO.
MONTH DAY YEAR 05000 NUMBER oa II FACILITY NAME DOCKET NUMBER 12 13 2014 2014
- - 011 00 2015 05000
- 9. OEPRATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 3 D 20.2201(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(i)(C)
D 50.73(a)(2)(vii)
D 20.2201(d)
D 20.2203(a)(3)(ii)
~ 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
- 10. POWER LEVEL D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(ii)
D 50.36(c)(1)(ii)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x) 000 D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4)
D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(C) 0 OTHER D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(v)(D)
Specify in Abstract below or in
CAUSE OF EVENT
- 3. PAGE 3 of 4 A failure modes analysis was performed and concluded that a definitive cause of the crack could not be determined without destructively testing the failed weld which was not possible; however, the most likely cause was determined to stem from a poor weld provided by the manufacturer of the pump component.
Based on a metallurgical visual examination and lSI observations during the weld repair, the apparent cause for the vendor weld failure was from a lack of fusion between weld passes.
ANALYSIS/SAFETY SIGNIFICANCE
The actual safety consequence was a violation of the Unit 1 Technical Specification, Section 3.4.4, "Reactor Coolant System (RCS)". The RCS leakage shall be limited to no pressure boundary leakage.
The potential consequence was a % inch unisolable pipe leak from the Reactor Coolant Pressure Boundary.
Based on review of the Unit 1 Technical Specification Bases (Section 3.4.4 ):
The allowable Reactor Coolant System (RCS) operational leakage limits are based on the predicted and experimentally observed behavior of pipe cracks. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage were also considered. The evidence from experiments suggests that, for leakage even greater than the specified unidentified leakage limits, the probability is small that the imperfection or crack associated with such leakage would grow rapidly.
The unidentified leakage flow limit allows time for corrective action before the Reactor Coolant Pressure Boundary (RCPB) could be significantly compromised. The five gallons per minute (gpm) limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Crack behavior from experimental programs shows that leakage rates of hundreds of gallons per minute will precede crack instability During the December 13, 2014 entry of the drywell, identified leakage from the 1 B RXR Lower Seal Cavity pipe connection was characterized as a mist.
The consequences are considered minimal given the small size of the leak.
CORRECTIVE ACTIONS
A weld repair on the cracked weld was performed.
Pipe-to-union welds on replacement shaft connections currently installed on the 1 P401 Band 2P401A RXR pumps will be cut out and re-welded. These actions are planned for the Unit 1 refueling outage in 2016 and the Unit 2 refueling outage in 2017.
COMPONENT FAILURE INFORMATION
The failed weld was a pipe-to-union weld on the Unit 1 "B" RXR Pump Lower Seal Cavity Vent piping. The Replacement RXR pump shafts and related components were procured from FLOWSERVE (formerly BW/IP) in the mid-1990s. The replacement shafts were stored in the warehouse until installed (one on the 2A RXR pump in 2013 and the other on the 1A RXR pump in 2014). The weld that ultimately failed was a vendor weld that existed at the time of procurement. The apparent cause was a poor weld provided by the manufacturer of the pump component due to a lack of fusion between weld passes.
PREVIOUS SIMILAR EVENTS
LER 2012-007-01: "Unplanned Shutdown Due to Elevated Drywell Unidentified Leakage," dated November 20, 2012. Although this was also reported as a degraded condition as a result of pressure boundary leakage, the cause was due to cyclic fatigue.
Various condition reports were identified that involved cracked/leaking welds; however, the causes of the identified cracks/leaks were related to vibration or stress corrosion cracking.