05000289/LER-2017-002, Regarding Leak at High Pressure Connection on Reactor Coolant Pump a Thermal Barrier
| ML17038A405 | |
| Person / Time | |
|---|---|
| Site: | Crane  (DPR-050) |
| Issue date: | 02/06/2017 |
| From: | Haaf T Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TMI-17-009 LER 17-002-00 | |
| Download: ML17038A405 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(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)(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(b)(2)(ii) |
| 2892017002R00 - NRC Website | |
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~ Exelon Generation Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 February 6, 2017 TMl-17-009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMl-1)
RENEWED FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289
SUBJECT:
LICENSEE EVENT REPORT (LER) NO. 2017-002-00 10 CFR 50.73 "Leak At High Pressure Connection on Reactor Coolant Pump "A" Thermal Barrier" This report is being submitted in accordance with 10 CFR 50.73(a)(2)(ii)(A). For additional information regarding this LER, please contact Mike Fitzwater, Sr. Regulatory Engineer, TMI Unit 1 Regulatory Assurance at (717) 948-8228.
There are no regulatory commitments contained in this LER.
- - 2~
Thomas P. Haaf Plant Manager, Three Mile Island Unit 1 Exelon Generation Co., LLC cc:
Administrator, NRC Region I TMI Senior Resident Inspector TMl-1 Project Manager R. R. Janati, Chief, Division of Nuclear Safety, Pennsylvania Department of Environmental Protection, Bureau of Radiation Protection
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMS: NO. 3150-0104 EXPIRES: 10/31/2018 (06-2016)
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 e-mail (See Page 2 for required number of digits/characters for each block) to lofocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a (See NUREG-1022, R.3 for instruction and guidance for completing this form means used to impose an information collection does not display a currently valid OMB control httg://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Three Mile Island Unit 1 05000289 1 OF 5
- 4. TITLE Leak At High Pressure Connection on Reactor Coolant Pump "A" 1hermal Barrier
- 5. EVENT DATE
- 6. LEA NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOl,.VED I
SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 12 07 2016 2017 - 002
- - 00 02 06 2017 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (CheckaUthatapply)
D 20.2201 (b>
D 20.2203(a)(3)(i)
~ 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
N D 20.2201(d)
D 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1 >
D 20.2203(a)(4)
D so.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)
- 10. POWER LEVEL D 20.2203(a)(2Hii>
D 50.36(c)(1)(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.11 (a)(4)
D 20.2203(a)(2)(iii)
D so.36(c)(2>
D 50.73(a)(2)(v)(B)
D 13.11(a)(s>
D 20.2203(a)(2)(iv)
D so.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1>
0%
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)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
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- I D 50.73(a)(2)(i)(C)
D OTHER i
Specify in Abstract below or in j
'f Contributing Cause #2: The vibrations driven by RCP. operation had an amplified effect on the piping due to a RCP seal package modification made in 2015 reducing the margin between the natural frequency of the pressure taps and the driving frequencies of the pump.
An extent of condition review was performed. The extent of condition included visual inspections of the remaining labyrinth seal differential pressure (dP) low pressure tap on RC-P-1A and the labyrinth seal dP high and low pressure taps on RC-P-1 B and RC-P-1 D. RC-P-1 C was not included in the extent of condition exams as the new pump installed in 1999 did not have the labyrinth seal dP taps. The RC-P-1A labyrinth seal dP low pressure tap included both external and internal VT-1 examinations. Due to operational sequence of the RCPs and the associated higher vibrations on RC-P-1A relative to RC-P-18/C/D, the three remaining thermal barrier welded connections on RC-P-1A were inspected (seal injection/make-up location, the Intermediate Closed Cooling Water (ICCW) heat exchanger (HX) outlet, and the ICCW heat exchanger (HX) inlet). No indications were observed during these VT-1 examinations. The ICCW HX outleVinlet and Makeup seal injection connections are a different configuration, and were not modified by the seal package modification and are therefore less susceptible to vibration fatigue. As a result, those connections are not included in the root cause corrective action extent of condition.
During start-up following repairs, when the plant reached normal operating pressure and temperature, VT-2 examinations were performed on all labyrinth seal connections. No leakage was observed.
The most probable root cause of the RC-P-1A high side labyrinth seal dP tap leak was an existing weld defect from original construction that decreased the fatigue strength of the connection. Specifically, this weld defect created a stress concentration that allowed vibration fatigue to initiate/propagate a flaw. While all welds (socket or buttwelds) are susceptible to weld defects, socket welded connections are the most susceptible to fatigue failures (Contributing Cause #1 ). Additionally, high cycle fatigue (HCF) would be expected to cause failure within the first couple of operating cycles of a component, and older, unmodified connections are not susceptible (Contributing Cause #2). Therefore, this extent of cause can be limited to the RCP labyrinth seal dP taps by use of the contributing causes.
C.
ANALYSIS I SAFETY SIGNIFICANCE This event resulted in a nonisolable leak in the RCS pressure boundary. The leak was not a threat to the safety of the reactor, since it was already in a safe shutdown condition. The leak extended a planned maintenance shutdown that had an occupational radiological dose implication of approximately 3.579 Rem.
The actual safety consequence was minimal. The leak was approximately 0.5 gpm and the unit was in a shutdown condition at the time of discovery of the leak.
A leak at one of the remaining Reactor Coolant Pump labyrinth seal dP pressure taps was assessed.
Independent analysis supports that the failure likely would not have occurred due to vibrational effects alone. Although high amplitude vibrations experienced during startup are required to initiate a flaw, visual inspections after startup did not identify any leakage.
- 3. LER NUMBER SEQUENTIAL NUMBER 002
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REV NO.
00 D.
CORRECTIVE ACTIONS
Immediate action was taken to remove the RC-P-1A labyrinth seal differential pressure (dP) high pressure tap. The welded pipe/flange connection was completely removed and a welded plug was installed to seal this connection in the RC-P-1 A thermal barrier.
The extent of condition will be addressed by implementing a design change to eliminate the vulnerability of a similar failure occurring on the socket welds for the remaining labyrinth seal dP taps on the three applicable RCPs (five locations) during the Fall 2017 refueling and maintenance outage.
E.
PREVIOUS OCCURENCES
Previous Events LEA 1995-003-00 Reactor Coolant Leak Due To A Cracked Weld In A Cold Leg Drain Line LER 2012-003-00 PRESSURIZER HEATER BUNDLE LEAK LER 2013-001-00 REACTOR COOLANT"B" COLD LEG DRAIN LINE FLAW LER 2013-001-01 REACTOR COOLANT"B" COLD LEG DRAIN LINE FLAW
- 2. DOCKET NUMBER YEAR 05000-289 2017 -
represent a reduction in the public health and safety.
- 3. LEA NUMBER SEQUENTIAL NUMBER 002 A previous PWSCC condition was reported in LER 50-289/2003-003-00.
This LER was submitted pursuant to 10 CFR 50.73(a)(2)(i)(A).
REV NO.
00 On November 7, 2013 TMl-1 was in a refueling shutdown mode for the planned T1 R20 refueling and maintenance outage. During a planned ISi volumetric examination of the reactor coolant '~B" cold leg drain line a flaw in the pipe weld was discovered. The flaw was located in a 2-inch drain line elbow to pipe weld.
The flaw was determined to not meet acceptance standards under ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws", and the RCS strength boundary was considered degraded. This condition required reporting under 1 O CFR 50.72(b)(3)(ii)(A) as a non-emergency degraded condition. The eight hour report was made at 13:02 on November 07, 2013 documented under EN 49512. An extent of condition and ISi scope expansion was performed. Similar pipe configurations were examined and their structural integrity to meet ASME code requirements was confirmed. The flawed section of "B" cold leg drain line was cut out and replaced. This LER was supplemented after receipt of the destructive laboratory test results *of the flawed section. There was no actual breach of the RCS that resulted in leakage. There were no adverse safety consequences or safety implications that resulted from this event and this event did not affect the health and safety of the public.
LER 2013-001-01 Supplement submitted June 20, 2014.
A destructive laboratory examination and finite element analysis (FEA) was performed on the removed pipe section. The root cause of the crack is unknown but believed to be the result of geometry induced focusing of lower levels of stress, not capable of inducing cracking alone, but when combined at a geometric feature such as a lack of fusion (LOF) at a natural notch, initiated the crack. The lower enerav requirements of propaQation then Qoverned the crack growth.
- Energy Industry Identification System (EllS), System Identification (SI) and Component Function Identification (CFI) Codes are included in brackets, [Sl/CFI] where applicable, as required by.1 O CFR 50.73 (b)(2)(ii)(F). Page _5_ of _5_