05000387/LER-2015-009, Regarding Pressure Boundary Leakage from an Inadequate Weld Repair in Small Bore Pump Seal Vent Piping
| ML16011A415 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/11/2016 |
| From: | Franke J Susquehanna, Talen Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-7419 LER 15-009-00 | |
| Download: ML16011A415 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3872015009R00 - NRC Website | |
text
JAN 1 1 2016 Jon A. Franke Susquehanna Nuclear, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 Jon.Franke@TalenEnergy.com U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387/2015-009-00 UNIT 1 LICENSE NO. NPF-14 PLA-7419 TALEN~
ENERGY 10 CFR 50.73 Docket No. 50-387 Attached is Licensee Event Report (LER) 50-387/2015-009-00. The LER reports a condition concerning Reactor Coolant Pressure Boundary leakage. This condition was determined to be repmiable in accordance with 10 CFR 50.73(a)(2)(ii)(A) and 10 CFR 50.73(a)(2)(i)(B), as a condition resulting in both a principal safety barrier degradation and being prohibited by Technical Specifications.
There were no actual consequences to the health and safety of the public as a result of this event.
s no new regulatory commitments.
Attachment: LER 50-3 87/2015-009-00 Copy:
NRC Region I Mr. J. E. Greives, NRC Sr. Resident Inspector Ms. T. E. Hood, NRC Project Manager Mr. M. Shields, P A DEP/BRP
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
~-PAGE Susquehanna Steam Electric Station Unit 1 05000387 1 OF 4
- 4. TITLE Pressure Boundary Leakage From an Inadequate Weld Repair in Small Bore Pump Seal Vent Piping
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR
/SEQUENTIAL/
REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 11 13 2015 2015
- - 009
- - 00 01
\\ \\
2016 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.2201(b)
D 20.2203(a)(3)(i)
~ 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 4 D 20.2201(d)
D 20.2203(a)(3)(ii)
D 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)
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)
- 10. POWER LEVEL 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) 1:8] 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) 0 OTHER Specify in Abstract below or in
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR I SEQUENTIAL I REV NUMBER NO.
05000387 2015
- - 009
- - 00 November 13, 2015 Drywell leakage is identified again to be at the 1 P401 B Reactor Recirculation Pump Seal Connection No. 8.
November 21, 2015 A weld repair was again performed on the 1 B reactor recirculation pump seal.
This condition was reported on November 13, 2015, at 0022 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, in accordance with 10 CFR 50.72(b)(3)(ii)(A) as a principal safety barrier degradation (EN 51538). This License Event Report (LER) is written in accordance with 10 CFR 50.73(a)(ii)(A) and 10 CFR 50.73(a)(2)(i)(B) as a condition that resulted in a principal safety barrier degradation with evidence of reactor coolant pressure boundary leakage, which is a condition prohibited by Technical Specifications.
CAUSE OF THE EVENT
The cause of this event is a decision making process that did not consider the technical merits of prudent versus allowable repair options. More specifically, the previous December 2014 weld repair did not fully excavate the weld and remove the J-groove, and thereby eliminate the presence of the crack.
A lack of fusion between original vendor weld passes, or at the root of the weld at the interface with the remaining vendor weld, is also considered to be a causal factor, contributing to further weld cracking and resulting reactor coolant pressure boundary leakage. This lack of fusion remaining within in the vendor weld, not in the December 2014 weld repair itself, created an initiating site for fatigue cracking to worsen.
ANALYSIS/SAFETY SIGNIFICANCE
The consequence of this condition was a violation of the Unit 1 Technical Specification (TS), Section 3.4.4, "Reactor Coolant System (RCS)," a reportable condition. The TS requires RCS leakage be limited to no pressure boundary leakage.
The safety significance of this condition is minimal. A 3/4-inch unisolable pipe leak from the reactor coolant pressure boundary from this condition is unlikely, given the drywell leakage observed prior to this repair, e.g., 0.6 GPM. Allowable reactor coolant system operational leakage limits are based on the predicted and experimentally observed behavior of pipe cracks. The probability is small that the imperfection or crack associated with such leakage would grow rapidly. The unidentified leakage flow limit of Technical Specifications allows time for corrective action before the reactor coolant pressure boundary could be significantly compromised. The five 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. In summary, given the size of the leak, there were no actual consequences to the health and safety of the public.
NRC FORM 366 (1 1-201 5)
CORRECTIVE ACTIONS
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR I SEQUENTIAL I REV NUMBER NO.
05000387 2015
. 009
. 00 Completed-Repair of the cracked weld was performed prior to the restart of Unit 1.
Planned - The procedure for control of welding will be revised to include additional requirements for addressing repair of cracked welds on small bore reactor coolant pressure boundary piping, such that the cognizant weld engineer consults with the Non-Destructive Examination (NDE) supervision, and/or manager of programs engineering to:
identify possible repair options with respect to removal of the weld metal defects (e.g., excavate unacceptable defects to sound metal, remove entire weld to base metal, etc.),
assess risk for each identified repair option, (e.g., potential consequences if the method is ineffective),
document selected repair option, and ensure the planner assigned to plan a weld repair work document is provided a specific repair option that documents their instruction.
PREVIOUS SIMILAR EVENTS
This event is similar to the following event, involving a weld crack repair and leakage at this same location.
LER 50-387/2014-011-00: "Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by an Inadequate Weld," issued February 11, 2015.
NRC FORM 366 (11-2015)
JAN 1 1 2016 Jon A. Franke Susquehanna Nuclear, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 Jon.Franke@TalenEnergy.com U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387/2015-009-00 UNIT 1 LICENSE NO. NPF-14 PLA-7419 TALEN~
ENERGY 10 CFR 50.73 Docket No. 50-387 Attached is Licensee Event Report (LER) 50-387/2015-009-00. The LER reports a condition concerning Reactor Coolant Pressure Boundary leakage. This condition was determined to be repmiable in accordance with 10 CFR 50.73(a)(2)(ii)(A) and 10 CFR 50.73(a)(2)(i)(B), as a condition resulting in both a principal safety barrier degradation and being prohibited by Technical Specifications.
There were no actual consequences to the health and safety of the public as a result of this event.
s no new regulatory commitments.
Attachment: LER 50-3 87/2015-009-00 Copy:
NRC Region I Mr. J. E. Greives, NRC Sr. Resident Inspector Ms. T. E. Hood, NRC Project Manager Mr. M. Shields, P A DEP/BRP
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
~-PAGE Susquehanna Steam Electric Station Unit 1 05000387 1 OF 4
- 4. TITLE Pressure Boundary Leakage From an Inadequate Weld Repair in Small Bore Pump Seal Vent Piping
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR
/SEQUENTIAL/
REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 11 13 2015 2015
- - 009
- - 00 01
\\ \\
2016 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.2201(b)
D 20.2203(a)(3)(i)
~ 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 4 D 20.2201(d)
D 20.2203(a)(3)(ii)
D 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)
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)
- 10. POWER LEVEL 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) 1:8] 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) 0 OTHER Specify in Abstract below or in CONDITIONS PRIOR TO THE EVENT Unit 1 - Mode 4, Unplanned shutdown, 0 Percent Other than the leaking weld itself there were no systems, structures, or components at the start of the event that contributed to the event. Prior to this condition on November 12, 2015, at 1132 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.30726e-4 months <br />, the Unit 1 reactor automatically scrammed due to one Main Steam Isolation Valve (MSIV) unanticipated closure causing a High Pressure Reactor Protection System trip. During the unplanned forced outage, a drywell entry and drywellleakage investigation were performed.
EVENT DESCRIPTION
On November 13, at 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br /> during drywell entry, a leak was reported on the "B" Reactor Recirculation (RXR) [EllS System Identifier: AD] Pump [EllS Component Identifier: P] Lower Seal Cavity Vent piping [EllS Component Identifier: PSP]. The leak was identified at the inboard pipe-to-union weld and required a weld repair prior to returning Unit 1 to service. The leaking weld was attached to the vendor supplied replacement pump seal chamber. The affected piping weld is for 3/4-inch piping, schedule 80 SA-479 TP304 or TP316, union 3000# SA-182 Gr F304 or F316. The affected piping had been in service for approximately 11 months following a previous repair of the weld at this location during the December 2014 forced outage of Unit 1, (e.g., LER 2014-011-00, issued February 11, 2015). A timeline of relevant events follows:
Mid-1990s Apri12013 April 2014 December 2014 December 2014 through 2015 Replacement RXR pump shafts and related components were procured from FLOWSERVE (formerly BW/IP). The replacement shafts were stored in the warehouse until installed (one on the 2A RXR pump in 2013 and the other on the 1 B RXR pump in 2014). The weld discovered to have cracking and leakage in December 2014 was a vendor weld that existed at the time of procurement.
The 2A RXR pump shaft and related components were replaced with components procured form FLOWSERVE in the mid-1990s.
The 1 B RXR Pump shaft and related components were replaced with components procured from FLOWSERVE in the mid-1990s.
Unit 1 initiation of a plant shutdown due to unidentified drywell leakage, discovery of weld cracking at this location, and its repair.
Unit 1 drywellleakage took a step increase from 0.1 to 0.2 gallons per minute (GPM) on December 28, 2014. Subsequently, gradual increases in excess drywell leakage were observed, (e.g., 0.2 to 0.6 GPM).
November 12, 2015 Unit 1 reactor scram due to 'B' inboard MSIV closure during surveillance testing, (unrelated to this condition).
NRC FORM 366 (1 1-2015)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR I SEQUENTIAL I REV NUMBER NO.
05000387 2015
- - 009
- - 00 November 13, 2015 Drywell leakage is identified again to be at the 1 P401 B Reactor Recirculation Pump Seal Connection No. 8.
November 21, 2015 A weld repair was again performed on the 1 B reactor recirculation pump seal.
This condition was reported on November 13, 2015, at 0022 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, in accordance with 10 CFR 50.72(b)(3)(ii)(A) as a principal safety barrier degradation (EN 51538). This License Event Report (LER) is written in accordance with 10 CFR 50.73(a)(ii)(A) and 10 CFR 50.73(a)(2)(i)(B) as a condition that resulted in a principal safety barrier degradation with evidence of reactor coolant pressure boundary leakage, which is a condition prohibited by Technical Specifications.
CAUSE OF THE EVENT
The cause of this event is a decision making process that did not consider the technical merits of prudent versus allowable repair options. More specifically, the previous December 2014 weld repair did not fully excavate the weld and remove the J-groove, and thereby eliminate the presence of the crack.
A lack of fusion between original vendor weld passes, or at the root of the weld at the interface with the remaining vendor weld, is also considered to be a causal factor, contributing to further weld cracking and resulting reactor coolant pressure boundary leakage. This lack of fusion remaining within in the vendor weld, not in the December 2014 weld repair itself, created an initiating site for fatigue cracking to worsen.
ANALYSIS/SAFETY SIGNIFICANCE
The consequence of this condition was a violation of the Unit 1 Technical Specification (TS), Section 3.4.4, "Reactor Coolant System (RCS)," a reportable condition. The TS requires RCS leakage be limited to no pressure boundary leakage.
The safety significance of this condition is minimal. A 3/4-inch unisolable pipe leak from the reactor coolant pressure boundary from this condition is unlikely, given the drywell leakage observed prior to this repair, e.g., 0.6 GPM. Allowable reactor coolant system operational leakage limits are based on the predicted and experimentally observed behavior of pipe cracks. The probability is small that the imperfection or crack associated with such leakage would grow rapidly. The unidentified leakage flow limit of Technical Specifications allows time for corrective action before the reactor coolant pressure boundary could be significantly compromised. The five 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. In summary, given the size of the leak, there were no actual consequences to the health and safety of the public.
NRC FORM 366 (1 1-201 5)
CORRECTIVE ACTIONS
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 2. DOCKET NUMBER
- 3. LER NUMBER YEAR I SEQUENTIAL I REV NUMBER NO.
05000387 2015
. 009
. 00 Completed-Repair of the cracked weld was performed prior to the restart of Unit 1.
Planned - The procedure for control of welding will be revised to include additional requirements for addressing repair of cracked welds on small bore reactor coolant pressure boundary piping, such that the cognizant weld engineer consults with the Non-Destructive Examination (NDE) supervision, and/or manager of programs engineering to:
identify possible repair options with respect to removal of the weld metal defects (e.g., excavate unacceptable defects to sound metal, remove entire weld to base metal, etc.),
assess risk for each identified repair option, (e.g., potential consequences if the method is ineffective),
document selected repair option, and ensure the planner assigned to plan a weld repair work document is provided a specific repair option that documents their instruction.
PREVIOUS SIMILAR EVENTS
This event is similar to the following event, involving a weld crack repair and leakage at this same location.
LER 50-387/2014-011-00: "Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by an Inadequate Weld," issued February 11, 2015.
NRC FORM 366 (11-2015)