05000388/LER-2015-004, Regarding Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by Vibration and Stiff Pipe Connection

Regarding Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by Vibration and Stiff Pipe Connection

ML15161A008
Person / Time
Site:	Susquehanna
Issue date:	06/10/2015
From:	Franke J Susquehanna, Talen Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7337 LER 15-004-00
Download: ML15161A008 (5)

LER-2015-004, Regarding Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by Vibration and Stiff Pipe Connection

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)(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
3882015004R00 - NRC Website

text

• JUN 1 0 2015 Jon A. Franke Site Vice President U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 Jon. Franke@talenenergy.com SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-388/2015-004-00 UNIT 2 LICENSE NO. NPF -22 PLA-7337 TALEN~

ENERGY 10 CFR 50.73 Docket No. 50-388 Attached is Licensee Event Report (LER) 50-388/2015-004-00. 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). This event is also being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.

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.

~~

~ ::T h=P?.e~

J. A. Franke Attachment: LER 388/2015-004-00 Copy:

NRC Region I Mr. J. E. Greives, NRC Sr. Resident Inspector Mr. J. A. Whited, NRC Project Manager Mr. B. R. Fuller, PA DEP/BRP

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017 (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 />.

,~.. ->~**"'~

r~~\\

Reported lessons learned are incorporated into the licensing process and fed back to industry.

~"- ':f)

LICENSEE EVENT REPORT (LER)

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections

.... 1 ~ **....

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-10202, (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 OMS 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 2 05000388 1 of4

4. TITLE Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by Vibration and Stiff Pipe Connection

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 Olo FACILITY NAME DOCKET NUMBER 04 11 2015 2015

- 004 00

\\0 2015 05000

9. OEPRATlNG MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 3 D 2o.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) 1:8:1 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)

D 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) 1:8:1 50.73(a)(2)(i)(B)

D 50.73(a)(2)(v)(D)

Specify in Abstract below or in

CAUSE OF EVENT

The direct cause of the weld crack was determined to be high cycle, low amplitude vibration fatigue.

The apparent cause involved the combined effects of an unrecognized system vibrational mode and a stiff pipe connection to the seal flange. The vibrational mode was introduced in 2013 during a replacement of the motor, shaft, and seal that included piping changes to accommodate an elevation change. The stiff pipe configuration was introduced during piping and pipe support modifications made in 1999 and 2000.

Causal factors included the presence of a stress concentrator (minor lack of fusion) on the root of the weld and two early life cavitation events in 1983. The minor lack of fusion is a fairly common occurrence on this type of weld and does not, by itself, render the weld defective.

ANALYSIS/SAFETY SIGNIFICANCE

The small leak was a violation of the Unit 2 Technical Specification, Section (TS) 3.4.4, "Reactor Coolant System (RCS)." The RCS leakage shall be limited to no pressure boundary leakage.

The condition was identified after the unit was shutdown. During plant operation, unidentified drywell leakage was measured at approximately 0.25 gallons per minute (gpm). The TS allowable is 5 gpm. The weld failure was a significant contributor to the overall unidentified drywell leakage. There was no evidence available to plant operators at the time to substantiate that leakage was from the reactor coolant pressure boundary. The unit was shut down for a scheduled refuel and inspection outage. During containment walk-downs, the subject condition was identified and classified as pressure boundary leakage.

The potential consequence was a % inch unisolable pipe leak from the Reactor Coolant Pressure Boundary.

Based on review of the Unit 2 TS Bases (Section 3.4.4 ):

The allowable 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 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 April 11, 2015 entry of the drywell, identified leakage from the 2A RXR Seal piping weld was described as a spray out of the top of the pipe in a fan pattern as opposed to a steady stream.

CORRECTIVE ACTIONS

U.S. NUCLEAR REGULATORY COMMISSION

6. LER NUMBER

3. PAGE I

SEQUENTIAL I REVISION NUMBER NUMBER 4 of 4

- 004

- 00

1. Unit 2: The piping configurations for Pipe Connection #2 and #6 on 2P401A and for Pipe Connection #1 and #2 on 2P401 B were modified to add flexibility in the pipe run between the seal flange and first support and minimize vibration modes coincident with known pump resonance frequencies 1X, 2X, and 5X.

2. Unit 1: The stiffened connections on 1 P401A and 1 P401 B (Connection #1 for each pump) are not reactor coolant pressure boundaries. The piping configurations for Pipe Connection #1 on 1 P401 A and on 1 P401 B will be modified to add flexibility in the pipe run between the seal flange and first support and minimize vibration modes coincident with known pump resonance frequencies 1X, 2X, and 5X.

COMPONENT FAILURE INFORMATION

The failed component was a seal flange weld associated with the pressure and vent piping to the upper seal chamber (Connection #2) for the Unit 2 RXR pump. The apparent cause involved the combined effects of an unrecognized system vibration and a stiff pipe connection to the seal flange.

PREVIOUS SIMILAR EVENTS

LER 50-387/2014-011-00: "Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by Inadequate Weld," dated February 11, 2015. Although this was also reported as a degraded condition as a result of pressure boundary leakage, the cause was due an inadequate weld.

LER 50-387/2012-007-01: "Unplanned Shutdown Due to Elevated Drywell Unidentified Leakage," dated November 20, 2012.

Susquehanna identified eight condition reports (CRs) involving cracked welds on small bore piping. Seven out of the eight CRs identify vibration as the cause of the cracked weld, and one CR identified lack of weld fusion as the cause of the cracked weld. The seven CRs identifying socket weld failures occurred between 1992 and 2004.

Susquehanna completed an initiative to mitigate the effects of system vibration on the fatigue life of socket welds inside containment. Volumetric inspections are performed on the socket welds considered to be most vulnerable to vibration-induced fatigue. The goal of the inspections is to ensure that the socket welds are of good quality with no latent cracks. If the weld is of good quality then the existing weld is built up with a full circumference EPRI 2x1 socket weld overlay. If the weld is not of good quality then the entire weld is removed and a full circumference EPRI 2x1 socket weld is installed. Previous corrective actions appear to be effective given that no non-vendor weld failures have occurred since the 2004 time frame.