05000387/LER-2012-007, Regarding Unplanned Shutdown Due to Unidentified Drywell Leakage

Regarding Unplanned Shutdown Due to Unidentified Drywell Leakage

ML12230A169
Person / Time
Site:	Susquehanna
Issue date:	08/17/2012
From:	Helsel J Susquehanna
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
PLA-6905 LER 12-007-00
Download: ML12230A169 (5)

LER-2012-007, Regarding Unplanned Shutdown Due to Unidentified Drywell Leakage

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 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)(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)(B), Prohibited by Technical Specifications
3872012007R00 - NRC Website

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I Jeffrey M. Helsel Nuclear Plant Manager PPLSusquehanna,LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3510 Fax 570.542.1504 jmhelsel@pplweb.com

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U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387/2012-007-00 UNIT 1 LICENSE NO. NPF-14 PLA-6905 Docket No 50-387 Attached is Licensee Event Report (LER) 50-387/2012-007-00, submitted for the Susquehanna Unit 1 unplanned shutdown on June 19, 2012.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(i)(A) for a Technical Specification (TS) required shutdown and 10 CFR 50.73(a)(2)(ii)(A) for a degraded condition.

There were no actual consequences to the health and safety of the public as a result -of this event and there are no regulatory commitments associated with this LER.

Please direct any questions regarding this report to Mr. Duane L. Filchner at (610)774-7819.

Attachment: LER 50-387/2012-007-00 cc:

NRC Region I Mr. P. W. Finney, NRC Sr. Resident Inspector Ms. C. J. Sanders, NRC Project Manager Mr. L. J. Winker, DEP/BRP TM

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-01 04 EXPIRES:1 0/31/2013 (10-2010)

, the NRC digits/characters for each block) may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Susquehanna Steam Electric Station (SSES) Unit 1 05000387 1 OF 4

4. TITLE Unplanned Shutdown due to Unidentified Drywell Leakage

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 NUMBER NO.

MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 06 19 2012 2012

- 007

- 00 u<l 17 2012 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 D 2o.22o1(b)

D 20.2203(a)(3)(i)

D 50.73(a)(2)(i)(C)

D 50.73(a)(2)(vii)

D 20.22o1 (d)

D 20.2203(a)(3)(ii)

[8:1 50. 73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203(a)(1)

D 20.2203(a)(4)

D 50.73(a)(2)(ii)(B)

D 50i73(a)(2)(viii)(B)

10. POWER LEVEL 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) 100%

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)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

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

D 73.71(a)(4) 0 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)

[8:1 50. 73(a)(2)(i)(A) 0 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

ANALYSIS/SAFETY SIGNIFICANCE

Actual Consequences:

U.S. NUCLEAR REGULATORY COMMISSION

6. LER NUMBER

3. PAGE SEQUENTIAL REVISION NUMBER NUMBER 30F4

- 007

- 00 The actual safety consequences of this event are minimal. An orderly shutdown of the Unit 1 reactor commenced when unidentified leakage reached approximately 1.1 gpm. The shutdown was performed well before the unidentified drywell leak rate reached the Technical Specification (TS) 3.4.4 limit of 5 gpm.

Potential Consequences:

The potential safety consequence of this event is that, if not corrected, this leak could have progressed to the point of reaching critical crack size. Reaching the critical crack size would have resulted in a PRA initiating event such as a scram or mid sized loss of coolant accident (LOCA). This could have resulted in an emergency plan entry for excessive drywellleakage and may have resulted in reactor coolant system loss through a 4 inch pipe. However, this potential consequence is unlikely to have occurred. There is significant margin between the 5 gpm maximum unidentified drywellleakage allowed by TS 3.4.4 and the leakage expected prior to reaching the critical'crack size.

Upon completion of the root cause evaluation, the actual safety significance and any potential consequences will be included in a supplement to this LER.

CORRECTIVE ACTIONS TAKEN An immediate corrective action taken was to modify the chemical decontamination assembly connected to the reactor recirculation piping. This consisted of a redesigned 4 inch diameter chemical decontamination flange connection on both the "1A" and "1 8" reactor recirculation pump suction lines. The length of the 4 inch diameter pipe was reduced from 6 inches to approximately 3.5 inches to produce a new configuration that is not susceptible to the cyclic fatigue caused by reactor recirculation pump vane passing frequency that is very close to the natural frequency of the assembly. This modification also eliminated the IGSCC that were identified during metallurgical examinations.

Extent of condition inspections were performed to provide assurance that similar reactor recirculation and reactor water cleanup system piping has not been similarly affected. The inspections were completed, no vibration related issues were identified, and no additional fatigue flaws were identified in the expanded scope population.

An in-service leak test was conducted in accordance with the requirements of ASME Section XI at a test pressure of 1035 psig to ensure pressure boundary integrity of the modified decontamination flange connections.

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As left impact resonance tests were performed on the "1A" and "1 8" reactor recirculation suction pipe decontamination flanges utilizing a test acceptance criteria for natural frequency greater than or equal to 150 Hz.

As a compensatory action for SSES Unit 2, the Unit 2 reactor recirculation pump speed is limited to 1600 rpm to provide margin to the resonant frequency of the reactor recirculation decontamination connection assemblies.

CORRECTIVE ACTIONS PLANNED U.S. NUCLEAR REGULATORY COMMISSION

6. LERNUMBER

3. PAGE SEQUENTIAL REVISION NUMBER NUMBER 40F4

- 007

- 00 Completion of the critical flaw size analysis to determine the continued growth crack size at which application of design basis loads would have resulted in failure.

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Additional corrective actions determined during completion of the root cause evaluation will be included in a supplement to this LER.

PREVIOUS SIMILAR EVENTS

There are no Susquehanna LER's related to TS required shutdown due to unidentified drywellleakage or reactor recirculation system leakage.

Industry Operating Experience (OE20348) for a similar event at Hope Creek LER 354/2005-002-00 was reviewed to ensure an understanding of the causes and actions taken. The investigation, inspection, and repair activities performed by SSES are consistent with those performed by Hope Creek.